Omniclassical Diffusion in Low Aspect Ratio Tokamaks
International Nuclear Information System (INIS)
Mynick, H.E.; White, R.B.; Gates, D.A.
2004-01-01
Recently reported numerical results for axisymmetric devices with low aspect ratio A found radial transport enhanced over the expected neoclassical value by a factor of 2 to 3. In this paper, we provide an explanation for this enhancement. Transport theory in toroidal devices usually assumes large A, and that the ratio B p /B t of the poloidal to the toroidal magnetic field is small. These assumptions result in transport which, in the low collision limit, is dominated by banana orbits, giving the largest collisionless excursion of a particle from an initial flux surface. However in a small aspect ratio device one may have B p /B t ∼ 1, and the gyroradius may be larger than the banana excursion. Here, we develop an approximate analytic transport theory valid for devices with arbitrary A. For low A, we find that the enhanced transport, referred to as omniclassical, is a combination of neoclassical and properly generalized classical effects, which become dominant in the low-A, B p /B t ∼ 1 regime. Good agreement of the analytic theory with numerical simulations is obtained
Current drive and profile control in low aspect ratio tokamaks
International Nuclear Information System (INIS)
Chan, V.S.; Chiu, S.C.; Lin-Liu, Y.R.; Miller, R.L.; Turnbull, A.D.
1995-07-01
The key to the theoretically predicted high performance of a low aspect ratio tokamak (LAT) is its ability to operate at very large plasma current*I p . The plasma current at low aspect ratios follows the approximate formula: I p ∼ (5a 2 B t /Rqψ) [(1 + κ 2 )/2] [A/(A - 1)] where A quadruple-bond R/a which was derived from equilibrium studies. For constant qψ and B t , I p can increase by an order of magnitude over the case of tokamaks with A approx-gt 2.5. The large current results in a significantly enhanced β t (quadruple-bond β N I p /aB t ) possibly of order unity. It also compensates for the reduction in A to maintain the same confinement performance assuming the confinement time τ follows the generic form ∼ HI p P -1 / 2 R 3 / 2 κ 1 / 2 . The initiation and maintenance of such a large current is therefore a key issue for LATs
Plasma features and alpha particle transport in low-aspect ratio tokamak reactor
International Nuclear Information System (INIS)
Xu Qiang; Wang Shaojie
1997-06-01
The results of the experiment and theory from low-aspect ratio tokamak devices have proved that the MHD stability will be improved. Based on present plasma physics and extrapolation to reduced aspect ratio, the feature of physics of low-aspect ratio tokamak reactor is discussed primarily. Alpha particle confinement and loss in the self-justified low-aspect ratio tokamak reactor parameters and the effect of alpha particle confinement and loss for different aspect ratio are calculated. The results provide a reference for the feasible research of compact tokamak reactor. (9 refs., 2 figs., 3 tabs.)
Development path of low aspect ratio tokamak power plants
International Nuclear Information System (INIS)
Stambaugh, R.D.; Chan, V.S.; Miller, R.L.
1997-03-01
Recent advances in tokamak physics indicate the spherical tokamak may offer a magnetic fusion development path that can be started with a small size pilot plant and progress smoothly to larger power plants. Full calculations of stability to kink and ballooning modes show the possibility of greater than 50% beta toroidal with the normalized beta as high as 10 and fully aligned 100% bootstrap current. Such beta values coupled with 2--3 T toroidal fields imply a pilot plant about the size of the present DIII-D tokamak could produce ∼ 800 MW thermal, 160 MW net electric, and would have a ratio of gross electric power over recirculating power (Q PLANT ) of 1.9. The high beta values in the ST mean that E x B shear stabilization of turbulence should be 10 times more effective in the ST than in present tokamaks, implying that the required high quality of confinement needed to support such high beta values will be obtained. The anticipated beta values are so high that the allowable neutron flux at the blanket sets the device size, not the physics constraints. The ST has a favorable size scaling so that at 2--3 times the pilot plant size the Q PLANT rises to 4--5, an economic range and 4 GW thermal power plants result. Current drive power requirements for 10% of the plasma current are consistent with the plant efficiencies quoted. The unshielded copper centerpost should have an adequate lifetime against nuclear transmutation induced resistance change and the low voltage, high current power supplies needed for the 12 turn TF coil appear reasonable. The favorable size scaling of the ST and the high beta mean that in large sizes, if the copper TF coil is replaced with a superconducting TF coil and a shield, the advanced fuel D-He 3 could be burned in a device with Q PLANT ∼ 4
All Metal Iron Core For A Low Aspect Ratio Tokamak
International Nuclear Information System (INIS)
Gates, D.A.; Jun, C.; Zatz, I.; Zolfaghari, A.
2010-01-01
A novel concept for incorporating a iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance separators between the toroidal field windings. This design avoids the inherent problems of a multiturn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron-induced conductivity. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance separators. Useful loop voltage time histories have been obtained using achievable resistivities.
Formation and sustainment of a low aspect ratio tokamak by a series of plasma injections
International Nuclear Information System (INIS)
Shimamura, Shin; Taniguchi, Makoto; Takahashi, Tsutomu; Nogi, Yasuyuki
1995-01-01
A low aspect ratio tokamak plasma was generated and sustained by injecting a series of plasmas from a magnetized coaxial gun into a flux conserver with toroidal field. The magnetized coaxial gun was supplied by an oscillating current with a d.c. component. The first few current pulses injected plasma and helicity into the flux conserver. This pulse helicity injection method worked effectively to maintain the low aspect ratio tokamak. 8 refs., 5 figs
Wave-driver options for low-aspect-ratio steady-state tokamak reactors
International Nuclear Information System (INIS)
Ehst, D.A.
1981-02-01
Low aspect ratio designs are proposed for steady-state tokamak reactors. Benefits stem from reduced major radius and lessened stresses in the toroidal field coils, resulting in possible cost savings in the tokamak construction. In addition, a low aspect ratio (A = 2.6) permits the application of a bundle divertor capable of diverting 3-T fields to a power reactor using STARFIRE technology. Such a low aspect ratio is possible with the elimination of poloidal field coils in the central hole of the tokamak, which implies a need for noninductive current drive. Several plasma waves are considered for this application, and it appears likely that a candidate can be found which reduces the electric power for current maintenance to an acceptable value
Characterization of peeling modes in a low aspect ratio tokamak
Bongard, M. W.; Thome, K. E.; Barr, J. L.; Burke, M. G.; Fonck, R. J.; Hinson, E. T.; Redd, A. J.; Schlossberg, D. J.
2014-11-01
Peeling modes are observed at the plasma edge in the Pegasus Toroidal Experiment under conditions of high edge current density (Jedge ˜ 0.1 MA m-2) and low magnetic field (B ˜ 0.1 T) present at near-unity aspect ratio. Their macroscopic properties are measured using external Mirnov coil arrays, Langmuir probes and high-speed visible imaging. The modest edge parameters and short pulse lengths of Pegasus discharges permit direct measurement of the internal magnetic field structure with an insertable array of Hall-effect sensors, providing the current profile and its temporal evolution. Peeling modes generate coherent, edge-localized electromagnetic activity with low toroidal mode numbers n ⩽ 3 and high poloidal mode numbers, in agreement with theoretical expectations of a low-n external kink structure. Coherent MHD fluctuation amplitudes are found to be strongly dependent on the experimentally measured Jedge/B peeling instability drive, consistent with theory. Peeling modes nonlinearly generate ELM-like, field-aligned filamentary structures that detach from the edge and propagate radially outward. The KFIT equilibrium code is extended with an Akima spline profile parameterization and an improved model for induced toroidal wall current estimation to obtain a reconstruction during peeling activity with its current profile constrained by internal Hall measurements. It is used to test the analytic peeling stability criterion and numerically evaluate ideal MHD stability. Both approaches predict instability, in agreement with experiment, with the latter identifying an unstable external kink.
A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks
International Nuclear Information System (INIS)
Woolley, R.D.
2009-01-01
A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress
A Novel Demountable TF Joint Design for Low Aspect Ratio Spherical Torus Tokamaks
International Nuclear Information System (INIS)
Woolley, Robert D.
2009-01-01
A novel shaped design for the radial conductors and demountable electrical joints connecting inner and outer legs of copper TF system conductors in low aspect ratio tokamaks is described and analysis results are presented. Specially shaped designs can optimize profiles of electrical current density, magnetic force, heating, and mechanical stress.
Non-inductive current drive via helicity injection by Alfven waves in low aspects ratio Tokamak
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
1996-01-01
A theoretical investigation of radio frequency (RF) current drive via helicity injection in low aspect ratio tokamaks was carried out. A current-carrying cylindrical plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell was considered. Toroidal features of low aspect ratio tokamaks were simulated by incorporation of the following effects: (i) arbitrarily small aspect ratio, R o /a ≡ 1/ε (ii) strongly sheared equilibrium magnetic field; and (iii) relatively large poloidal component of the equilibrium magnetic field. The study concentrates on the Alfven continuum, i.e. the case in which the wave frequency satisfies the condition {ω Alf (r)} min ≤ω≥{ω Alf (r)} max , where ω Alf (r)≡ω[n(r),B o (o)] is an eigenfrequency of the shear Alfven wave (SAW). Thus, using low-p, ideal magneto-hydrodynamics, the wave equation with correct boundary (matching) conditions was solved, the RF field components were found and subsequently, current drive , power deposition and efficiency were computed. The results of our investigation clearly demonstrate the possibility of generation of RF-driven currents via helicity injection by Alfven waves in low aspect ratio tokamaks, in the SAW mode. A special algorithm was developed which enables the selection of the antenna parameters providing optimal current drive efficiency. (authors)
Resistive vs. total power depositions by Alfven modes in pre-heated low aspect ratio tokamaks
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
2004-01-01
The power deposition of fast waves launched by a LFS located antenna in a pre-heated, strongly non-uniform low aspect ratio tokamak (START) is investigated. The rigorous computational results indicate a total power deposition by far larger than that predicted for Alfven continuum eigenmodes in cylindrical plasmas. For toroidal wave numbers |N| > 1, the resistive and total power depositions are almost equal. (author)
Finite Larmor radius effects on Alfven wave current drive in low-aspect ratio tokamaks
International Nuclear Information System (INIS)
Komoshvili, K.; Cuperman, S.; Bruma, C.
1998-01-01
Alfven wave current drive (AWCD) in low-aspect ratio (A≡R/a=1/ε > or approx. 1) tokamaks (LARTs) is studied numerically. For this, the full-wave equation (E parallel ≠0) with a Vlasov-based dielectric tensor is solved by relaxation techniques, subject to appropriate boundary conditions at the plasma centre and at the plasma-vacuum interface, as well as the concentric antenna current sheet and at the external metallic wall. A systematic investigation of the physical characteristics of the AWCD generated in LARTs when kinetic effects are considered is carried out and illustrative results are presented and discussed. (author)
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1998-01-01
Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high β, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E p arallel=0) MHD and non-ideal, resistive (E p arallel≠0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E p arallel≠0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (author)
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Cuperman, S.; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, Tel Aviv (Israel)
1998-08-01
Several low aspect ratio (spherical) Tokamaks (ST's) are now in operation or under construction. These devices would permit cost-effective and attractive embodiment of future fusion reactors: they would provide high {beta}, good confinement and steady state operation at modest field values. Now, a steady state reactor has to be sustained by non-inductively driven currents. Recently, the generation of non-inductive current drive by Alfven waves (AWCD) has been investigated theoretically within the framework of ideal (E{sub p}arallel=0) MHD and non-ideal, resistive (E{sub p}arallel{ne}0) MHD; however, in all these cases, the tokamak device consisted of a cylindrical plasma with simulated toroidal effects. Rather encouraging results have been obtained. In this work we further investigate AWCD in ST's as follows: (i) we use consistent equilibrium profiles with neoclassical conductivity corresponding to an ohmic START discharge; (ii) incorporate effects due to neoclassical conductivity in the elements of the resistive MHD dielectric tensor, in the solution of the full (E{sub p}arallel{ne}0) wave equation as well as in the calculation of AWCD; and (iii) carry out a systematic search for antenna parameters optimizing the AWCD. (author)
Stable equilibria for bootstrap-current-driven low aspect ratio tokamaks
International Nuclear Information System (INIS)
Miller, R.L.; Lin-Liu, Y.R.; Turnbull, A.D.; Chan, V.S.; Pearlstein, L.D.; Sauter, O.; Villard, L.
1997-01-01
Low aspect ratio tokamaks (LATs) can potentially provide a high ratio of plasma pressure to magnetic pressure β and high plasma current I at a modest size. This opens up the possibility of a high-power density compact fusion power plant. For the concept to be economically feasible, bootstrap current must be a major component of the plasma current, which requires operating at high β p . A high value of the Troyon factor β N and strong shaping is required to allow simultaneous operation at a high-β and high bootstrap fraction. Ideal magnetohydrodynamic stability of a range of equilibria at aspect ratio 1.4 is systematically explored by varying the pressure profile and shape. The pressure and current profiles are constrained in such a way as to assure complete bootstrap current alignment. Both β N and β are defined in terms of the vacuum toroidal field. Equilibria with β N ≥8 and β∼35%endash 55% exist that are stable to n=∞ ballooning modes. The highest β case is shown to be stable to n=0,1,2,3 kink modes with a conducting wall. copyright 1997 American Institute of Physics
Collisional Transport in a Low Aspect Ratio Tokamak -- Beyond the Drift Kinetic Formalism
International Nuclear Information System (INIS)
Gates, D.A.; White, R.B.
2004-01-01
Calculations of collisional thermal and particle diffusivities in toroidal magnetic plasma confinement devices order the toroidal gyroradius to be small relative to the poloidal gyroradius. This ordering is central to what is usually referred to as neoclassical transport theory. This ordering is incorrect at low aspect ratio, where it can often be the case that the toroidal gyroradius is larger than the poloidal gyroradius. We calculate the correction to the particle and thermal diffusivities at low aspect ratio by comparing the diffusivities as determined by a full orbit code (which we refer to as omni-classical diffusion) with those from a gyroaveraged orbit code (neoclassical diffusion). In typical low aspect ratio devices the omni-classical diffusion can be up to 2.5 times the calculated neoclassical value. We discuss the implications of this work on the analysis of collisional transport in low aspect ratio magnetic confinement experiments
Non-inductive current drive via helicity injection by Alfven waves in low-aspect-ratio tokamaks
Energy Technology Data Exchange (ETDEWEB)
Cuperman, S.; Bruma, C.; Komoshvili, K. [Tel Aviv Univ. (Israel). Sackler Faculty of Exact Sciences
1996-08-01
A theoretical investigation of radio-frequency (RF) current drive via helicity injection in low aspect ratio tokamaks is carried out. A current-carrying cylindrical plasma surrounded by a helical sheet-current antenna and situated inside a perfectly conducting shell is considered. Toroidal features of low-aspect-ratio tokamaks are simulated by incorporating the following effects: (i) arbitrarily small aspect ratio, R{sub O}/a ``identical to`` 1/{epsilon}; (ii) strongly sheared equilibrium magnetic field; and (iii) relatively large poloidal component of the equilibrium magnetic field. This study concentrates on the Alfven continuum, i.e. the case in which the wave frequency satisfies the condition {l_brace}{omega}{sub Alf}({tau}){r_brace}{sub min}{r_brace} {<=} {omega} {<=} {l_brace}{omega}{sub Alf}({tau}){r_brace}{sub max}, where {omega}{sub Alf}({tau}) ``identical to`` {omega}{sub Alf}[n({tau}), B{sub O}({tau})] is an eigenfrequency of the shear Alfven wave (SAW). Thus, using low-{beta} magnetohydrodynamics, the wave equation with correct boundary (matching) conditions is solved, the RF field components are found, and subsequently current drive, power deposition and efficiency are computed. The results of our investigation clearly demonstrate the possibility of generation of RF-driven currents via helicity injection by Alfven waves in low-aspect-ratio tokamaks, in the SAW mode. A special algorithm is developed that enables one to select the antenna parameters providing optimal current drive efficiency. (Author).
Study of an optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak
Energy Technology Data Exchange (ETDEWEB)
Hong, Bong Guen, E-mail: bghong@jbnu.ac.kr [Department of Quantum System Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of); Kim, Hoseok [Department of Applied Plasma Engineering, Chonbuk National University, 567 Baekje-daero, Jeonju, Jeonbuk 54896 (Korea, Republic of)
2016-11-15
Highlights: • Optimum configuration of a transmutation reactor based on a low aspect ratio tokamak was found. • Inboard and outboard radial build are determined by plasma physics, engineering and neutronics constraints. • Radial build and equilibrium fuel cycle play a major role in determining the transmutation characteristics. - Abstract: We determine the optimal configuration of a transmutation reactor based on a low-aspect-ratio tokamak. For self-consistent determination of the radial build of the reactor components, we couple a tokamak systems analysis with a radiation transport calculation. The inboard radial build of the reactor components is obtained from plasma physics and engineering constraints, while outboard radial builds are mainly determined by constraints on neutron multiplication, the tritium-breeding ratio, and the power density. We show that the breeding blanket model has an effect on the radial build of a transmutation blanket. A burn cycle has to be determined to keep the fast neutron fluence plasma-facing material below its radiation damage limit. We show that the radial build of the transmutation reactor components and the equilibrium fuel cycle play a major role in determining the transmutation characteristics.
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Kommoshvili, K [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Cuperman, S [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Bruma, C [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel)
2003-03-01
Kinetic effects in the conversion of fast waves to Alfven waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvenic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxiliary energy source for the successful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects.
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
International Nuclear Information System (INIS)
Kommoshvili, K; Cuperman, S; Bruma, C
2003-01-01
Kinetic effects in the conversion of fast waves to Alfven waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvenic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxiliary energy source for the successful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
Kommoshvili, K.; Cuperman, S.; Bruma, C.
2003-03-01
Kinetic effects in the conversion of fast waves to Alfvèn waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvènic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxilliary energy source for the succesful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects.
Technological and environmental prospects of low aspect ratio tokamak reactor VECTOR
International Nuclear Information System (INIS)
Nishio, S.; Tobita, K.; Isono, T.; Nakamura, Y.; Sato, M.; Sakurai, S.; Yamauchi, M.; Nishitani, T.; Tani, K.; Sengoku, S.; Kudo, Y.; Song, Y.; Tokimatsu, K.; Shinya, K.; Senda, I.; Konishi, S.
2005-01-01
Optimization of VECTOR design parameters has led to a fusion output of 2.5 GW with a small reactor weight of 8,800 tons. Current ramp-up simulation with TSC (Tokamak Simulation Code) demonstrated a stable current ramp-up with externally non-inductive current drive and bootstrap current when a central current is induced externally enough to avoid an extreme reversed shear such as current hole. It was numerically confirmed that such a low-A reactor would have an advantage of α-particle confinement. The divertor is designed to use a flux expansion of scrape-off-layer in the divertor region to maintain the heat flux on the divertor plate to be lower than 10 MW/m 2 without remote radiative cooling. Because of its compactness, CO 2 emission in a life cycle of a VECTOR power plant is estimated to be as low as 3.2 g-CO 2 /kWh, being lower than that of an ITER-sized DEMO reactor (4.9 g-CO 2 /kWh). As to the waste management of VECTOR, on the basis of reactor design and radiological considerations, we suggest reusing a liquid metal breeding material (PbLi) and neutron shield material (TiH 2 ) in successive reactors. Due to this waste management, its disposal waste would be reduced to as low as 3,000-4,000 tons, which is comparable with the radioactive waste of a light water reactor (4,000 tons in metal). (author)
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.
2003-01-01
We investigate quantitatively the possibility of turbulence suppression through the generation of transport barriers in pre-heated low aspect ratio tokamaks (LARTs) by the sheared electric fields generated by externally driven rf waves in the frequency range ω A ≅ ci (ω A and ω ci are, respectively, Alfven and ion cyclotron frequencies). To this aim the following sequential steps are followed: (1) Solutions of the resistive two-fluid model full wave equation for a realistic LART configuration (D-shape cross-section and aspect ratio, R/a ∼> 1; as well as suitably located low field side, LFS, antenna) upon using a quite general dielectric tensor operator; (2) Calculation of the ponderomotive forces and their magnetic surface averages; (3) Solution of a strongly non-linear differential equation for the quasi-stationary radial electric field, including the particle orbit squeezing effects, based on the results of steps (1) and (2); and (4) Calculation of the radial flow shear, S perpend ; for both banana and potato collisional regimes
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Cuperman, S. [Tel Aviv Univ. (Israel). School of Physics and Astronomy; Komoshvili, K. [The College of Judea and Samaria, Ariel (Israel)
2003-05-01
We investigate quantitatively the possibility of turbulence suppression through the generation of transport barriers in pre-heated low aspect ratio tokamaks (LARTs) by the sheared electric fields generated by externally driven rf waves in the frequency range {omega}{sub A} {approx_equal} <{omega}{sub ci} ({omega}{sub A} and {omega}{sub ci} are, respectively, Alfven and ion cyclotron frequencies). To this aim the following sequential steps are followed: (1) Solutions of the resistive two-fluid model full wave equation for a realistic LART configuration (D-shape cross-section and aspect ratio, R/a {approx}> 1; as well as suitably located low field side, LFS, antenna) upon using a quite general dielectric tensor operator; (2) Calculation of the ponderomotive forces and their magnetic surface averages; (3) Solution of a strongly non-linear differential equation for the quasi-stationary radial electric field, including the particle orbit squeezing effects, based on the results of steps (1) and (2); and (4) Calculation of the radial flow shear, S{sub perpend}; for both banana and potato collisional regimes.
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
1999-01-01
The authors developed a consistent formalism for the full wave equation, appropriate for the study of propagation, absorption and wave conversion of externally launched waves in strongly toroidal, spherical tokamaks with non-circular cross-section. This includes also the formulation of rigorous regularity, boundary, gauge and periodicity conditions suitable for the exact solution of the wave equation for such devices
Burke, M. G.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Perry, J. M.; Reusch, J. A.; Schlossberg, D. J.
2017-07-01
Plasmas in the Pegasus spherical tokamak are initiated and grown by the non-solenoidal local helicity injection (LHI) current drive technique. The LHI system consists of three adjacent electron current sources that inject multiple helical current filaments that can reconnect with each other. Anomalously high impurity ion temperatures are observed during LHI with T i,OV ⩽ 650 eV, which is in contrast to T i,OV ⩽ 70 eV from Ohmic heating alone. Spatial profiles of T i,OV indicate an edge localized heating source, with T i,OV ~ 650 eV near the outboard major radius of the injectors and dropping to ~150 eV near the plasma magnetic axis. Experiments without a background tokamak plasma indicate the ion heating results from magnetic reconnection between adjacent injected current filaments. In these experiments, the HeII T i perpendicular to the magnetic field is found to scale with the reconnecting field strength, local density, and guide field, while {{T}\\text{i,\\parallel}} experiences little change, in agreement with two-fluid reconnection theory. This ion heating is not expected to significantly impact the LHI plasma performance in Pegasus, as it does not contribute significantly to the electron heating. However, estimates of the power transfer to the bulk ion are quite large, and thus LHI current drive provides an auxiliary ion heating mechanism to the tokamak plasma.
Energy Technology Data Exchange (ETDEWEB)
Cuperman, S. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel); Bruma, C. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel) and College of Judea and Samaria, 44837 Ariel (Israel)]. E-mail: edycb@post.tau.ac.il; Komoshvili, K. [School of Physics and Astronomy, Tel Aviv University, 69978 Tel-Aviv (Israel); College of Judea and Samaria, 44837 Ariel (Israel)
2007-03-05
The combined plasma non-uniformity effects on the energy deposition of Alfven waves launched by an external antenna in pre-heated spherical tokamaks are investigated. The following relevant physical processes are here possible: (a) the emergence of gaps in the shear Alfven continuum spectrum and the generation of discrete global Alfven eigenmodes with frequencies inside the gaps; (b) multi-wave interactions, interactions of gaps of the same kind (e.g., toroidicity induced) and of different kinds (toroidicity, ellipticity and triangularity induced) as well as of secondary order gaps arising when a pair of modes is coupled to one or more modes through other coupling parameters; (c) basic wave-plasma interactions as propagation, reflection, mode-conversion, tunneling and deposition. Thus, we solved numerically the full 2D wave equations for the vector and scalar potentials, using a quite general two-fluid resistive tensor-operator, without any geometrical limitations. The results obtained indicate the existence of antenna-launched wave characteristics for which the power is most efficiently coupled in outer regions of plasmas, which is of special interest for low aspect ratio tokamaks, e.g., for the generation of non-inductive current drive as well as for turbulence suppression and transport barriers formation.
International Nuclear Information System (INIS)
Bruma, C; Cuperman, S; Komoshvili, K
2003-01-01
In an effort to optimize the internal transport barriers (ITBs) generated by externally launched mode-converted fast waves (FWs) in pre-heated spherical tokamaks (STs), we have carried out a systematic parametric investigation with respect to the rf waves and antenna characteristics; as a study case, a START-like device has been considered. Within the framework of a plasma model including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped and passing electrons and ions, and starting with the solution of the full wave equation for a ST-plasma, we show that optimized ITBs, suitable for the stabilization of plasma turbulence (e.g. overpassing the maximum growth rate of the ITG-instability, γ ITG ) in STs can be generated by the aid of externally launched FW and mode-converted to kinetic Alfven waves. This result holds in spite of the limiting trapped-particles associated squeezing factor S present in the non-linear equation for E r (via the viscosity coefficient μ θi ∝|S| 3/2 , S = S(dE r /dr))
Energy Technology Data Exchange (ETDEWEB)
Bruma, C [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Cuperman, S [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel); Komoshvili, K [School of Physics and Astronomy, Tel Aviv University, 69978 Tel Aviv (Israel)
2003-04-01
In an effort to optimize the internal transport barriers (ITBs) generated by externally launched mode-converted fast waves (FWs) in pre-heated spherical tokamaks (STs), we have carried out a systematic parametric investigation with respect to the rf waves and antenna characteristics; as a study case, a START-like device has been considered. Within the framework of a plasma model including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped and passing electrons and ions, and starting with the solution of the full wave equation for a ST-plasma, we show that optimized ITBs, suitable for the stabilization of plasma turbulence (e.g. overpassing the maximum growth rate of the ITG-instability, {gamma}{sub ITG}) in STs can be generated by the aid of externally launched FW and mode-converted to kinetic Alfven waves. This result holds in spite of the limiting trapped-particles associated squeezing factor S present in the non-linear equation for E{sub r} (via the viscosity coefficient {mu}{sub {theta}}{sub i}{proportional_to}|S|{sup 3/2}, S = S(dE{sub r}/dr))
Energy Technology Data Exchange (ETDEWEB)
Cuperman, S.; Bruma, C.; Komoshvili, K
2003-05-12
The generation in low aspect ratio tokamaks (LARTs) of ponderomotive forces and non-inductive current drive by the resonant fast wave-plasma interaction with mode conversion to kinetic Alfven waves (KAWs) and subsequent deposition, mainly by resonant electron Landau damping, is considered. The calculations follow the rigorous solution of the full wave equations upon using a dielectric tensor operator consisting of (i) a parallel conductivity including both kinetic effects (collisionless Landau damping on passing electrons) and collisional damping on both trapped electrons and passing electrons+ions and (ii) perpendicular components provided by the resistive two-fluid model equations. The fast waves are launched by an antenna located on the low field side and extending {+-}45 deg. about the equatorial plane. A parametric investigation of the structure and importance of the various components of the ponderomotive forces and current drive generated in START-like plasmas is carried out and their suitability for supplementing the required non-rf toroidal equilibrium current is demonstrated.
Flutter analysis of low aspect ratio wings
Parnell, L. A.
1986-01-01
Several very low aspect ratio flat plate wing configurations are analyzed for their aerodynamic instability (flutter) characteristics. All of the wings investigated are delta planforms with clipped tips, made of aluminum alloy plate and cantilevered from the supporting vehicle body. Results of both subsonic and supersonic NASTRAN aeroelastic analyses as well as those from another version of the program implementing the supersonic linearized aerodynamic theory are presented. Results are selectively compared with the experimental data; however, supersonic predictions of the Mach Box method in NASTRAN are found to be erratic and erroneous, requiring the use of a separate program.
Low Aspect-Ratio Wings for Wing-Ships
DEFF Research Database (Denmark)
Filippone, Antonino; Selig, M.
1998-01-01
Flying on ground poses technical and aerodynamical challenges. The requirements for compactness, efficiency, manouverability, off-design operation,open new areas of investigations in the fieldof aerodynamic analysis and design. A review ofthe characteristics of low-aspect ratio wings, in- and out...
International Nuclear Information System (INIS)
Jarboe, T.R.; Nelson, B.A.
1995-01-01
This is the fifth Progress Report on the Helicity Injected Tokamak (HIT) at the University of Washington, Seattle, DOE Grant DE-FE06-90ER54095. This report covers the period of December 1, 1993 through December 31
Large Aspect Ratio Tokamak Study
International Nuclear Information System (INIS)
Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Wiseman, G.W.
1980-06-01
The Large Aspect Ratio Tokamak Study (LARTS) at Oak Ridge National Laboratory (ORNL) investigated the potential for producing a viable longburn tokamak reactor by enhancing the volt-second capability of the ohmic heating transformer through the use of high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were assessed in the context of extended burn operation. Using a one-dimensional transport code plasma startup and burn parameters were addressed. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the startup and shutdown portions of the tokamak cycle. A representative large aspect ratio tokamak with an aspect ratio of 8 was found to achieve a burn time of 3.5 h at capital cost only approx. 25% greater than that of a moderate aspect ratio design tokamak
Influence of external toroidal flux on low-aspect-ratio toroidal plasma
International Nuclear Information System (INIS)
Ikuno, S.; Natori, M.; Kamitani, A.
1999-01-01
In the HIST device, the external flux is generated by two kinds of currents: the current I s flowing along the symmetry axis and the bias coil current I D . The influence of the external flux on the MHD equilibrium and stability of the low-aspect-ratio toroidal plasma in the HIST device is investigated numerically. Equilibrium configurations of the low-aspect-ratio toroidal plasma in the HIST device are numerically determined by means of the combination of FDM and BEM. The influence of I s and I D on their stability is also investigated by using the Mercier criterion. The results of computations show that the Mercier limit decreases to zero with increasing I s and with decreasing I D . Moreover, either a further increase in I s or a further decrease in I D raises the Mercier limit considerably. Besides, the equilibrium configuration in the HIST device changes its state from spheromak through ultra-low q to tokamak with increasing I s and with decreasing I D . (author)
Oscillatory convection in low aspect ratio Czochralski melts
Anselmo, A.; Prasad, V.; Koziol, J.; Gupta, K. P.
1993-11-01
Modeling of the crucible in bulk crystal growth simulations as a right circular cylinder may be adequate for high aspect ratio melts but this may be unrealistic when the melt height is low. Low melt height is a unique feature of a solid feed continuous Czochralski growth process for silicon single crystals currently under investigation. At low melt heights, the crucible bottom curvature has a dampening effect on the buoyancy-induced oscillations, a source of inhomogeneities in the grown crystal. The numerical results demonstrate how the mode of convection changes from vertical wall-dominated recirculating flows to Benard convection as the aspect ratio is lowered. This phenomenon is strongly dependent on the boundary condition at the free surface of the melt, which has been generally considered to be either adiabatic or radiatively cooled. A comparison of the flow oscillations in crucibles with and without curved bottoms at aspect ratios in the range of 0.25 to 0.50, and at realistic Grashof numbers (10 7 < Gr < 10 8) illustrate that changing the shape of the crucible may be an effective means of suppressing oscillations and controlling the melt flow.
Configuration studies for a small-aspect-ratio tokamak stellarator hybrid
International Nuclear Information System (INIS)
Carreras, B.A.; Lynch, V.E.; Ware, A.
1996-08-01
The use of modulated toroidal coils offers a new path to the tokamak-stellarator hybrids. Low-aspect-ratio configurations can be found with robust vacuum flux surfaces and rotational transform close to the transform of a reverse-shear tokamak. These configurations have clear advantages in minimizing disruptions and their effect and in reducing tokamak current drive needs. They also allow the study of low-aspect-ratio effects on stellarator confinement in small devices
Large aspect ratio tokamak study
International Nuclear Information System (INIS)
Reid, R.L.; Holmes, J.A.; Houlberg, W.A.; Peng, Y.K.M.; Strickler, D.J.; Brown, T.G.; Sardella, C.; Wiseman, G.W.
1979-01-01
The Large Aspect Ratio Tokamak Study (LARTS) investigated the potential for producing a viable long burn tokamak reactor through enhanced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering, and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohmic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured
Design studies of low-aspect ratio quasi-omnigenous stellarators
International Nuclear Information System (INIS)
Spong, D.A.; Hirshman, S.; Whitson, J.C.
2001-01-01
Significant progress has been made in the development of new modest-size compact stellarator devices that could test optimization principles for the design of a more attractive reactor. These are 3 and 4 field period low-aspect-ratio quasi-omnigenous (QO) stellarators based on an optimization method that targets improved confinement, stability, ease of coil design, low-aspect-ratio, and low bootstrap current. (author)
Lee-side flow structures of very low aspect ratio cruciform wing–body configurations
CSIR Research Space (South Africa)
Tuling, S
2013-12-01
Full Text Available A numerical and experimental investigation was performed to study the dominant flow structures in the lee side of a cruciform wing–body configuration at supersonic speeds in the + orientation. The wings or strakes are of very low aspect ratio...
International Nuclear Information System (INIS)
Kim, Kyoung Jin; Peterson, Dennis R.
2008-01-01
Traditionally, the electrothermal gun design has the bore of a large aspect ratio: however, a low aspect ratio design with a shorter bore length has been employed for efficient production of metal plasma vapors and synthesis of nanomaterials. In a comparison of the arc resistance-current relationship, a low aspect ratio design is found to exhibit distinctively different characteristics compared to a high aspect ratio design, and this trend is explained by the scaling law of plasma properties including theory of plasma electrical conductivity. A one-dimensional isothermal model has been applied to the present experiments to confirm the scaling laws, and it was found that the present modification of the electrothermal gun is able to produce fully ionized metal plasma vapor, while the plasma vapor produced in a conventional design is partially ionized. Also, by reacting metal plasma vapors with the controlled gases in the reaction chamber, nanoscale materials such as aluminum oxide, aluminum nitride, and titanium oxide were synthesized successfully
Differential inertial focusing of particles in curved low-aspect-ratio microchannels
Energy Technology Data Exchange (ETDEWEB)
Russom, Aman; Gupta, Amit K; Nagrath, Sunitha; Di Carlo, Dino; Edd, Jon F; Toner, Mehmet [BioMEMS Resource Center, Center for Engineering in Medicine and Surgical Services, Massachusetts General Hospital, Shriners Hospital for Children, and Harvard Medical School, Boston, MA 02114 (United States)], E-mail: aman@kth.se
2009-07-15
Microfluidic-based manipulation of particles is of great interest due to the insight it provides into the physics of hydrodynamic forces. Here, we study a particle-size-dependent phenomenon based on differential inertial focusing that utilizes the flow characteristics of curved, low aspect ratio (channel width >> height), microfluidic channels. We report the emergence of two focusing points along the height of the channel (z-plane), where different sized particles are focused and ordered in evenly spaced trains at correspondingly different lateral positions within the channel cross-section. We applied the system for continuous ordering and separation of suspension particles.
Stability of low aspect ratio inverted flags and rods in a uniform flow
Huertas-Cerdeira, Cecilia; Sader, John E.; Gharib, Morteza
2016-11-01
Cantilevered elastic plates and rods in an inverted configuration, where the leading edge is free to move and the trailing edge is clamped, undergo complex dynamics when subjected to a uniform flow. The stability of low aspect ratio inverted plates and rods is theoretically examined, showing that it is markedly different from that of their large aspect ratio counterpart. In the limit of zero aspect ratio, the undeflected equilibrium position is found to be stable for all wind speeds. A saddle-node bifurcation emerges at finite wind speed, giving rise to a strongly deflected stable and a weakly deflected unstable equilibria. This theory is compared to experimental measurements, where good agreement is found. This research was supported by a Grant of the Gordon and Betty Moore Foundation, the Australian Research Council Grants scheme and a "la Caixa" Fellowship Grant for Post-Graduate Studies of "la Caixa" Banking Foundation.
Flutter analysis of hybrid metal-composite low aspect ratio trapezoidal wings in supersonic flow
Directory of Open Access Journals (Sweden)
Shokrollahi Saeed
2017-02-01
Full Text Available An effective 3D supersonic Mach box approach in combination with non-classical hybrid metal-composite plate theory has been used to investigate flutter boundaries of trapezoidal low aspect ratio wings. The wing structure is composed of two main components including aluminum material (in-board section and laminated composite material (out-board section. A global Ritz method is used with simple polynomials being employed as the trial functions. The most important objective of the present research is to study the effect of composite to metal proportion of hybrid wing structure on flutter boundaries in low supersonic regime. In addition, the effect of some important geometrical parameters such as sweep angle, taper ratio and aspect ratio on flutter boundaries were studied. The results obtained by present approach for special cases like pure metallic wings and results for high supersonic regime based on piston theory show a good agreement with those obtained by other investigators.
Origin of the outer layer of martian low-aspect ratio layered ejecta craters
Boyce, Joseph M.; Wilson, Lionel; Barlow, Nadine G.
2015-01-01
Low-aspect ratio layered ejecta (LARLE) craters are one of the most enigmatic types of martian layered ejecta craters. We propose that the extensive outer layer of these craters is produced through the same base surge mechanism as that which produced the base surge deposits generated by near-surface, buried nuclear and high-explosive detonations. However, the LARLE layers have higher aspect ratios compared with base surge deposits from explosion craters, a result of differences in thicknesses of these layers. This characteristics is probably caused by the addition of large amounts of small particles of dust and ice derived from climate-related mantles of snow, ice and dust in the areas where LARLE craters form. These deposits are likely to be quickly stabilized (order of a few days to a few years) from eolian erosion by formation of duricrust produced by diffusion of water vapor out of the deposits.
Hyperthermia in low aspect-ratio magnetic nanotubes for biomedical applications
Gutierrez-Guzman, D. F.; Lizardi, L. I.; Otálora, J. A.; Landeros, P.
2017-03-01
A simple model for the magnetization reversal process of low aspect-ratio ferromagnetic nanotubes (MNTs) is presented. Because of advantages over other geometries, these structures are interesting for biomedical applications, such as magnetic hyperthermia cancer therapy, where the heat released during magnetic reversal is used to destroy tumors. For example, the tubular geometry provides two independent functional surfaces that may be selectively manipulated and also gives a storage cavity. Owing to their large surface to weight ratio and low mass density, MNTs are not decanted by gravity. We calculated magnetic phase diagrams, energy barriers, nucleation fields, and the amount of dissipated heat and specific absorption rate for magnetite nanotubes. The geometrical parameters were varied, and simple formulae were used to optimize the tube response under alternating excitation, as required for magnetic hyperthermia applications.
International Nuclear Information System (INIS)
Shiina, S.; Yagi, Y.; Sugimoto, H.; Ashida, H.; Hirano, Y.; Koguchi, H.; Sakakita, H.; Taguchi, M.; Nagamine, Y.; Osanai, Y.; Saito, K.; Watanabe, M.; Aizawa, M.
2005-01-01
Dominant plasma self-induced current equilibrium is achieved together with the high β for the steady-state neoclassical reversed field pinch (RFP) equilibrium with low aspect ratio by broadening the plasma pressure profile. The RF-driven current, when the safety factor is smaller than unity, is much less than the self-induced current, which dominates (96%) the toroidal current. This neoclassical RFP equilibrium has strong magnetic shear or a high-stability beta (β t = 63%) due to its hollow current profile. It is shown that the obtained equilibrium is close to the relaxed-equilibrium state with a minimum energy, and is also robust against microinstabilities. These attractive features allow the economical design of compact steady-state fusion power plants with low cost of electricity (COE). (author)
Neoclassical viscosities in NCSX and QPS with few toroidal periods and low aspect ratios
International Nuclear Information System (INIS)
Nishimura, S.; Mikkelsen, D.R.; Ku, L.P.; Mynick, H.E.; Zarnstorff, M.C.; Spong, D.A.; Hirshman, S.P.
2008-01-01
Previously reported benchmarking examples for the analytical formulas of neoclassical viscosities were made implicitly assuming applications in a future integrated simulation system for the LHD (Large Helical Device). Therefore the toroidal period numbers assumed there were mainly N=10. In this kind of calculation, however, an implicit (or sometimes explicit) assumption of ι/N<<1 is sometimes included. This assumption is included not only in simplified bounce averaged drift kinetic equations for ripple diffusions, but also in the equation before the averaging for non-bounce-averaged effects determining neoclassical parallel viscosity and the banana-plateau diffusions. To clarify the applicability of the analytical methods even for configurations with extremely low toroidal period numbers (required for low aspect ratios), we show here recent benchmarking examples in NCSX (National Compact Stellarator Experiment) with N=3 and QPS (Quasi-poloidal Stellarator) with N=2. (author)
Theory and observation of compressional Alfven eigenmodes in low aspect ratio plasma
International Nuclear Information System (INIS)
Gorelenkov, N.N.
2002-01-01
A new theory of radially and poloidally localized Compressional Alfven Eigenmodes (CAE) in low aspect ratio plasma is reported. The theory is applied to identify recently observed instabilities in the MHz frequency range in National Spherical Torus experiments (NSTX). The frequency of observed CAEs is correlated with the characteristic Alfven velocity of the plasma. The observed high frequency modes are explained as CAEs driven by energetic beam ions. The CAE frequency is determined by the Alfven frequency at the mode location on the low field side of the plasma and is given approximately by ω CAE v A m=r, where m is the poloidal mode number, and r is the local minor radius. CAEs are destabilized by free energy in the energetic ion velocity space gradient via Doppler shifted cyclotron resonance with beam ions. Properties of the CAE instability driven by different NBI ion distributions are presented. (author)
International Nuclear Information System (INIS)
Onchi, T; Fujisawa, A; Sanpei, A; Himura, H; Masamune, S
2017-01-01
Permutation entropy and statistical complexity are measures for complex time series. The Bandt–Pompe methodology evaluates probability distribution using permutation. The method is robust and effective to quantify information of time series data. Statistical complexity is the product of Jensen–Shannon divergence and permutation entropy. These physical parameters are introduced to analyse time series of emission and magnetic fluctuations in low-aspect-ratio reversed-field pinch (RFP) plasma. The observed time-series data aggregates in a region of the plane, the so-called C – H plane, determined by entropy versus complexity. The C – H plane is a representation space used for distinguishing periodic, chaos, stochastic and noisy processes of time series data. The characteristics of the emissions and magnetic fluctuation change under different RFP-plasma conditions. The statistical complexities of soft x-ray emissions and magnetic fluctuations depend on the relationships between reversal and pinch parameters. (paper)
Physics issues for a very-low-aspect-ratio Quasi-Poloidal Stellarator (QPS)
International Nuclear Information System (INIS)
Lyon, J.F.; Berry, L.A.; Hirshman, S.P.
2003-01-01
A quasi-poloidal stellarator with very low plasma aspect ratio (R/a ∼ 2.7, 1/2-1/4 that of existing stellarators) is a new confinement approach that could ultimately lead to a high-beta compact stellarator reactor. The Quasi-Poloidal Stellarator (QPS) experiment is being developed to test key features of this approach. The QPS will study neoclassical and anomalous transport, stability limits at beta up to 2.5%, the configuration dependence of the bootstrap current, and equilibrium robustness. The quasi-poloidal symmetry leads to neoclassical transport that is much smaller than the anomalous transport. The reduced effective field ripple may also produce reduced poloidal viscosity, enhancing the ambipolar E x B poloidal drift and allowing larger poloidal flows for reduction of anomalous transport. A region of second stability exists in the QPS experiment at higher beta. Very-high-beta configurations with a tokamak-like transform profile have also been obtained with a bootstrap current 1/3-1/5 that in an equivalent tokamak. These configurations are stable to low-n ideal MHD kink and vertical instabilities for beta up to 11%. Ballooning-stable configurations are found for beta in the range 2% to 23%. (author)
A helium-cooled blanket design of the low aspect ratio reactor
International Nuclear Information System (INIS)
Wong, C.P.; Baxi, C.B.; Reis, E.E.; Cerbone, R.; Cheng, E.T.
1998-03-01
An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh
Confinement physic study in a small low-aspect-ratio helical device CHS
International Nuclear Information System (INIS)
Okamura, S.; Matsuoka, K.; Akiyama, R.
1999-01-01
The configuration parameter of the plasma position relative to the center of the helical coil winding is very effective one for controlling the MHD stability and the trapped particle confinement in Heliotron/Torsatron systems. But these two characteristics are contradictory to each other in this parameter. The inward shifted configuration is favorable for the drift-orbit-optimization but it is predicted unstable with the Mercier criterion. Various physics problems, such as electric field structure, plasma rotation and MHD phenomena, have been studied in CHS with a compromising intermediate position. With this standard configuration, CHS has supplied experimental results for understanding general toroidal confinement physics and low-aspect-ratio helical systems. In the recent experiments, it was found that the wide range of inward shifted configurations gives stable plasma discharges without any restriction to the special pressure profile. Such enhanced range of operation made it possible to study experimentally the drift-orbit-optimized configuration in the Heliotron/Torsatron systems. The effect of configuration improvement was studied with plasmas in a low collisionality regime. (author)
Barlow, Nadine G.; Boyce, Joseph M.; Cornwall, Carin
2014-09-01
Low-Aspect-Ratio Layered Ejecta (LARLE) craters are a unique landform found on Mars. LARLE craters are characterized by a crater and normal layered ejecta pattern surrounded by an extensive but thin outer deposit which terminates in a sinuous, almost flame-like morphology. We have conducted a survey to identify all LARLE craters ⩾1-km-diameter within the ±75° latitude zone and to determine their morphologic and morphometric characteristics. The survey reveals 140 LARLE craters, with the majority (91%) located poleward of 40°S and 35°N and all occurring within thick mantles of fine-grained deposits which are likely ice-rich. LARLE craters range in diameter from the cut-off limit of 1 km up to 12.2 km, with 83% being smaller than 5 km. The radius of the outer LARLE deposit displays a linear trend with the crater radius and is greatest at higher polar latitudes. The LARLE deposit ranges in length between 2.56 and 14.81 crater radii in average extent, with maximum length extending up to 21.4 crater radii. The LARLE layer is very sinuous, with lobateness values ranging between 1.45 and 4.35. LARLE craters display a number of characteristics in common with pedestal craters and we propose that pedestal craters are eroded versions of LARLE craters. The distribution and characteristics of the LARLE craters lead us to propose that impact excavation into ice-rich fine-grained deposits produces a dusty base surge cloud (like those produced by explosion craters) that deposits dust and ice particles to create the LARLE layers. Salts emplaced by upward migration of water through the LARLE deposit produce a surficial duricrust layer which protects the deposit from immediate removal by eolian processes.
Experiments on a low aspect ratio wing at low Reynolds numbers
Morse, Daniel R.
At the start of the 21st century much of the focus of aircraft design has been turned to unmanned aerial vehicles (UAVs) which generally operate at much lower speeds in higher risk areas than manned aircraft. One subset of UAVs are Micro Air Vehicles (MAVs) which usually are no larger than 20cm and rely on non-traditional shapes to generate lift at very low velocities. This purpose of this work is to describe, in detail with experimental methods, the flow field around a low aspect ratio wing operating at low Reynolds numbers and at high angles of attack. Quantitative measurements are obtained by Three Component Time Resolved Particle Image Velocimetry (3C TR PIV) which describe the mean and turbulent flow field. This research focuses on the leading edge separation zone and the vortex shedding process which occurs at the leading edge. Streamwise wing tip vortices which dominate the lift characteristics are described with flow visualization and 3C TR PIV measurements. Turbulent Kinetic Energy (TKE) is described at the leading edge over several angles of attack. Turbulent Reynolds stresses in all three directions are described over the wing span and several Reynolds numbers. Two primary cyclic processes are observed within the flow field; one low frequency oscillation in the separated region and one high frequency event associated with leading edge vortex formation and convection. Two length scales are proposed and are shown to match well with each other, one based on leading edge vortex shedding frequency and convective velocity and the other based on mean vortex separation distance. A new method of rendering velocity frequency content over large data sets is proposed and used to illustrate the different frequencies observed at the leading edge.
OCLATOR (One Coil Low Aspect Toroidal Reactor)
Energy Technology Data Exchange (ETDEWEB)
Yoshikawa, S.
1980-02-01
A new approach to construct a tokamak-type reactor(s) is presented. Basically the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which there will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak which should improve the ..beta.. limit, so that a low toroidal magnetic field strength might be acceptable, meaning that the NbTi superconducting wire could be used. This system is named OCLATOR (One Coil Low Aspect Toroidal Reactor).
Physics Issues in the Design of Low Aspect-Ratio, High-Beta, Quasi-Axisymmetric Stellarators
International Nuclear Information System (INIS)
Zarnstorff, M.C.; Berry, L.A.; Boozer, A.; Brooks, A.; Cooper, W.A.
2000-01-01
Compact stellarators have the potential to combine the best features of the stellarator and the advanced tokamak, offering steady state operation without current drive and potentially without disruptions at an aspect ratio similar to tokamaks. A quasi-axisymmetric stellarator is developed that is consistent with the boot-strap current and passively stable to the ballooning, kink, Mercier, vertical, and neoclassical tearing modes at b=4.1 % without need for conducting walls or external feedback. The configuration has good flux surfaces and fast ion confinement. Thermal transport analysis indicates that the confinement should be similar to tokamaks of the same size, allowing access to the b-limit with moderate power. Coils have been designed to reproduce the physics properties. Initial analysis indicates the coils have considerable flexibility to manipulate the configuration properties. Simulations of the current evolution indicate the kink-mode can remain stable during the approach to h igh-beta
International Nuclear Information System (INIS)
Vázquez, J; Rivera, M A; Hernando, J; Sánchez-Rojas, J L
2009-01-01
The response of commercial piezoelectric AFM probes for potential applications in the field of chemical or biological sensors operating in liquids is investigated using laser Doppler vibrometry. The present work investigates the roles played in the frequency response by the density and the viscosity of different water–glycerol mixtures, in a frequency range of up to 1 MHz in air. Since the width of the tested probes is relatively large (and hence the aspect ratio remains small), inertial loading effects dominate viscous effects, unlike in cantilevers characterized by larger aspect ratios. Measurements are compared with results provided by a simplified computer model of a probe immersed in an inviscid surrounding fluid
Kayatin, Matthew J.; Perry, Jay L.
2017-01-01
Traditional gas-phase trace contaminant control adsorption process flow is constrained as required to maintain high contaminant single-pass adsorption efficiency. Specifically, the bed superficial velocity is controlled to limit the adsorption mass-transfer zone length relative to the physical adsorption bed; this is aided by traditional high-aspect ratio bed design. Through operation in this manner, most contaminants, including those with relatively high potential energy are readily adsorbed. A consequence of this operational approach, however, is a limited available operational flow margin. By considering a paradigm shift in adsorption architecture design and operations, in which flows of high superficial velocity are treated by low-aspect ratio sorbent beds, the range of well-adsorbed contaminants becomes limited, but the process flow is increased such that contaminant leaks or emerging contaminants of interest may be effectively controlled. To this end, the high velocity, low aspect ratio (HVLA) adsorption process architecture was demonstrated against a trace contaminant load representative of the International Space Station atmosphere. Two HVLA concept packaging designs (linear flow and radial flow) were tested. The performance of each design was evaluated and compared against computer simulation. Utilizing the HVLA process, long and sustained control of heavy organic contaminants was demonstrated.
A modified atmospheric non-hydrostatic model on low aspect ratio grids
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Wen-Yih Sun
2012-04-01
Full Text Available It is popular to use a horizontal explicit and a vertical implicit (HE-VI scheme in the compressible non-hydrostatic (NH model. However, when the aspect ratio becomes small, a small time-interval is required in HE-VI, because the Courant-Fredrich-Lewy (CFL criterion is determined by the horizontal grid spacing. Furthermore, simulations from HE-VI can depart from the forward–backward (FB scheme in NH even when the time interval is less than the CFL criterion allowed. Hence, a modified non-hydrostatic (MNH model is proposed, in which the left-hand side of the continuity equation is multiplied by a parameter δ (4≤δ≤16, in this study. When the linearized MNH is solved by FB (can be other schemes, the eigenvalue shows that MNH can suppress the frequency of acoustic waves very effectively but does not have a significant impact on the gravity waves. Hence, MNH enables to use a longer time step than that allowed in the original NH. When the aspect ratio is small, MNH solved by FB can be more accurate and efficient than the NH solved by HE-VI. Therefore, MNH can be very useful to study cloud, Large Eddy Simulation (LES, turbulence, flow over complex terrains, etc., which require fine resolution in both horizontal and vertical directions.
A modified atmospheric non-hydrostatic model on low aspect ratio grids: part II
Directory of Open Access Journals (Sweden)
Wen-Yih Sun
2013-06-01
Full Text Available Sun et al. (2012 proposed a modified non-hydrostatic model (MNH, in which the left-hand side of the continuity equation is multiplied by a parameter δ (4≤δ≤16 in the article to suppress high-frequency acoustic waves. They showed that the MNH allows a longer time step than the original non-hydrostatic model (NH. The MNH is also more accurate and efficient than the horizontal explicit and vertical implicit scheme (HE–VI when the aspect ratio (Δx/Δz is small. In addition to multiplying a parameter δ, here we propose to add a smoothing on the right-hand side of the continuity equation in the MNH to damp shortest sound waves. Linear stability analysis and non-linear model simulations show that the MNH with smoothing (henceforth abbreviated as MNHS can use twice the time interval of the MNH while maintaining the same accuracy. The MNHS is also more accurate and efficient than HE–VI when the aspect ratio is small.
Engineering Design Study of Quasi-Axisymmetric Stellarator with Low Aspect Ratio
International Nuclear Information System (INIS)
Matsuoka, Keisuke; Okamura, Shoichi; Nishimura, Shin; Isobe, Mitsutaka; Suzuki, Chihiro; Shimizu, Akihiro; Tanaka, Nobuo; Hasegawa, Mitsuru; Naito, Hideji; Urata, Kazuhiro; Suzuki, Yutaka; Tsukamoto, Tadanori
2004-01-01
The engineering design of the quasi-axisymmetric stellarator CHS-qa is described, having a toroidal period number of 2, major radius of 1.5 m, and plasma aspect ratio of 3.2. Although the entire structure of the machine is highly nonaxisymmetric and deformative, the following major engineering concerns for the modular coils and the vacuum vessel have been resolved: (a) modular coil design (curvature and twist of conductors), (b) supporting structures for modular coils, (c) errors due to electromagnetic forces and misalignment in manufacturing processes (analysis shows that the magnetic surface is robust against such disturbances), (d) construction procedure for vacuum vessel and modular coils, and (e) ports for heating and diagnostics
International Nuclear Information System (INIS)
Iguchi, H.; Kubo, S.; Idei, H.
1993-01-01
Transport enhancement due to helical ripples is the main problem for a low-aspect-ratio helical system to survive as a magnetic fusion device. Optimization of the magnetic configuration has been experimentally studied for neutral beam heated plasmas in the Compact Helical System (CHS). A confinement regime compatible with the LHD scaling has been obtained by shifting the magnetic axis inward with respect to the minor axis of the helical windings. However a power balance analysis suggests that the improvement of plasma parameters has mainly been achieved by the improvement of power deposition. On the other hand, electron density profiles become peaked with the inward shifted magnetic axis in contrast to flattened profiles with the outward shifted one. A question arises: Does the magnetic structure really affect transport processes? In order to answer this question, it is most suitable to examine ECH plasmas in a low collisionality regime. In this paper we report some characteristics of the ECH plasmas in the low-aspect-ratio device CHS and discuss the effect of magnetic field ripples on transport processes. (author) 10 refs., 4 figs
Aspect Ratio Scaling of Ideal No-wall Stability Limits in High Bootstrap Fraction Tokamak Plasmas
International Nuclear Information System (INIS)
Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Maingi, R.; Sabbagh, S.A.; Soukhanovskii, V.; Stutman, D.
2003-01-01
Recent experiments in the low aspect ratio National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40 (2000) 557] have achieved normalized beta values twice the conventional tokamak limit at low internal inductance and with significant bootstrap current. These experimental results have motivated a computational re-examination of the plasma aspect ratio dependence of ideal no-wall magnetohydrodynamic stability limits. These calculations find that the profile-optimized no-wall stability limit in high bootstrap fraction regimes is well described by a nearly aspect ratio invariant normalized beta parameter utilizing the total magnetic field energy density inside the plasma. However, the scaling of normalized beta with internal inductance is found to be strongly aspect ratio dependent at sufficiently low aspect ratio. These calculations and detailed stability analyses of experimental equilibria indicate that the nonrotating plasma no-wall stability limit has been exceeded by as much as 30% in NSTX in a high bootstrap fraction regime
OCLATOR (One Coil Low Aspect Toroidal Reactor)
International Nuclear Information System (INIS)
Yoshikawa, S.
1980-02-01
A new approach to construct a tokamak-type reactor(s) is presented. Basically the return conductors of toroidal field coils are eliminated and the toroidal field coil is replaced by one single large coil, around which there will be placed several tokamaks or other toroidal devices. The elimination of return conductors should, in addition to other advantages, improve the accessibility and maintainability of the tokamaks and offer a possible alternative to the search for special materials to withstand large neutron wall loading, as the frequency of changeover would be increased due to minimum downtime. It also makes it possible to have a low aspect ratio tokamak which should improve the β limit, so that a low toroidal magnetic field strength might be acceptable, meaning that the NbTi superconducting wire could be used. This system is named OCLATOR
International Nuclear Information System (INIS)
Du, Ping; Lin, I-Kuan; Zhang, Xin; Lu, Hongbing
2010-01-01
Polydimethylsiloxane (PDMS)-based micropillars (or microcantilevers) have been used as bio-transducers for measuring cellular forces on the order of pN to µN. The measurement accuracy of these sensitive devices depends on appropriate modeling to convert the micropillar deformations into the corresponding reaction forces. The traditional approach to calculating the reaction force is based on the Euler beam theory with consideration of a linear elastic slender beam for the micropillar. However, the low aspect ratio in geometry of PDMS micropillars does not satisfy the slender beam requirement. Consequently, the Timoshenko beam theory, appropriate for a beam with a low aspect ratio, should be used. In addition, the inherently time-dependent behavior in PDMS has to be considered for accurate force conversion. In this paper, the Timoshenko beam theory, along with the consideration of viscoelastic behavior of PDMS, was used to model the mechanical response of micropillars. The viscoelastic behavior of PDMS was characterized by stress relaxation nanoindentation using a circular flat punch. A correction procedure was developed to determine the load–displacement relationship with consideration of ramp loading. The relaxation function was extracted and described by a generalized Maxwell model. The bending of rectangular micropillars was performed by a wedge indenter tip. The viscoelastic Timoshenko beam formula was used to calculate the mechanical response of the micropillar, and the results were compared with measurement data. The calculated reaction forces agreed well with the experimental data at three different loading rates. A parametric study was conducted to evaluate the accuracy of the viscoelastic Timoshenko beam model by comparing the reaction forces calculated from the elastic Euler beam, elastic Timoshenko beam and viscoelastic Euler beam models at various aspect ratios and loading rates. The extension of modeling from the elastic Euler beam theory to the
Shields, Matt
The development of Micro Aerial Vehicles has been hindered by the poor understanding of the aerodynamic loading and stability and control properties of the low Reynolds number regime in which the inherent low aspect ratio (LAR) wings operate. This thesis experimentally evaluates the static and damping aerodynamic stability derivatives to provide a complete aerodynamic model for canonical flat plate wings of aspect ratios near unity at Reynolds numbers under 1 x 105. This permits the complete functionality of the aerodynamic forces and moments to be expressed and the equations of motion to solved, thereby identifying the inherent stability properties of the wing. This provides a basis for characterizing the stability of full vehicles. The influence of the tip vortices during sideslip perturbations is found to induce a loading condition referred to as roll stall, a significant roll moment created by the spanwise induced velocity asymmetry related to the displacement of the vortex cores relative to the wing. Roll stall is manifested by a linearly increasing roll moment with low to moderate angles of attack and a subsequent stall event similar to a lift polar; this behavior is not experienced by conventional (high aspect ratio) wings. The resulting large magnitude of the roll stability derivative, Cl,beta and lack of roll damping, Cl ,rho, create significant modal responses of the lateral state variables; a linear model used to evaluate these modes is shown to accurately reflect the solution obtained by numerically integrating the nonlinear equations. An unstable Dutch roll mode dominates the behavior of the wing for small perturbations from equilibrium, and in the presence of angle of attack oscillations a previously unconsidered coupled mode, referred to as roll resonance, is seen develop and drive the bank angle? away from equilibrium. Roll resonance requires a linear time variant (LTV) model to capture the behavior of the bank angle, which is attributed to the
Second regime tokamak operation at large aspect ratio
International Nuclear Information System (INIS)
Navratil, G.A.
1989-01-01
This paper reviews the need for high beta in economic tokamak reactors and summarizes recent results on the scaling of the second regime beta limit for high-n ballooning modes using optimized pressure profiles as well as results on low-n mode stability at the first regime beta limit from the Columbia HBT tokamak. While several experiments have studied ballooning limits using high εβ p plasmas, the most important question for the use of the second stability regime for tokamak reactor improvement is how to achieve these high values of εβ p while at the same time increasing the value of beta to several times the Troyon beta limit. An approach to the study of this key question on beta limits using modest sized, large aspect ratio tokamaks is described. (author). 28 refs, 7 figs, 1 tab
Stability of high β large aspect ratio tokamaks
International Nuclear Information System (INIS)
Cowley, S.C.
1991-10-01
High β(β much-gt ε/q 2 ) large aspect ratio (ε much-gt 1) tokamak equilibria are shown to be always stable to ideal M.H.D. modes that are localized about a flux surface. Both the ballooning and interchange modes are shown to be stable. This work uses the analytic high β large aspect ratio tokamak equilibria developed by Cowley et.al., which are valid for arbitrary pressure and safety factor profiles. The stability results make no assumption about these profiles or the shape of the boundary. 14 refs., 4 figs
Graves, E. B.
1972-01-01
A study has been made to determine the aerodynamic characteristics of a low-aspect ratio cruciform missile model with all-movable wings and tails. The configuration was tested at Mach numbers from 1.50 to 4.63 with the wings in the vertical and horizontal planes and with the wings in a 45 deg roll plane with tails in line and interdigitated.
Stabe, Roy G.; Schwab, John R.
1991-01-01
A 0.767-scale model of a turbine stator designed for the core of a high-bypass-ratio aircraft engine was tested with uniform inlet conditions and with an inlet radial temperature profile simulating engine conditions. The principal measurements were radial and circumferential surveys of stator-exit total temperature, total pressure, and flow angle. The stator-exit flow field was also computed by using a three-dimensional Navier-Stokes solver. Other than temperature, there were no apparent differences in performance due to the inlet conditions. The computed results compared quite well with the experimental results.
Relation of vertical stability and aspect ratio in tokamaks
International Nuclear Information System (INIS)
Stambaugh, R.D.; Lao, L.L.; Lazarus, E.A.
1992-01-01
It is evaluated how the upper limit to plasma elongation κ, caused by vertical stability, varies with the aspect ratio A=R/a of the tokamak. Equilibria were generated with EFITD and the vertical stability was assessed by GATO. For a 'generic' tokamak with a superconducting wall conformal to the plasma shape and a distance 0.5 a away from the plasma edge and a constant current profile (q 0 =1.0, l i ≅1.0, q 95 =3.2) it is found that the maximum stable κ decreased only slowly from 2.65 at A=2.0 to 2.4 at A=6.0. To first order, a reasonable assumption in trade-off studies of new machine designs is no dependence of κ max on A. (author). Letter-to-the-editor. 13 refs, 3 figs, 1 tab
Stabe, R. G.; Whitney, W. J.; Moffitt, T. P.
1984-01-01
Experimental results are presented for a 0.767 scale model of the first stage of a two-stage turbine designed for a high by-pass ratio engine. The turbine was tested with both uniform inlet conditions and with an inlet radial temperature profile simulating engine conditions. The inlet temperature profile was essentially mixed-out in the rotor. There was also substantial underturning of the exit flow at the mean diameter. Both of these effects were attributed to strong secondary flows in the rotor blading. There were no significant differences in the stage performance with either inlet condition when differences in tip clearance were considered. Performance was very close to design intent in both cases. Previously announced in STAR as N84-24589
International Nuclear Information System (INIS)
Jarboe, T.R.; Nelson, B.A.
1992-01-01
In the paper we will detail the progress of the HIT experiment construction, including the following components: preliminary data and interpretation; diagnostic systems; vacuum vessel and pumping system; helicity source and power supplies; toroidal field coil and power supply; data acquisition system; collaboration with general atomics, with a brief summary given on each
Tight aspect ratio tokamak power reactor with superconducting TF coils
International Nuclear Information System (INIS)
Nishio, S.; Tobita, K.; Konishi, S.; Ando, T.; Hiroki, S.; Kuroda, T.; Yamauchi, M.; Azumi, M.; Nagata, M.
2003-01-01
Tight aspect ratio tokamak power reactor with super-conducting toroidal field (TF) coils has been proposed. A center solenoid coil system and an inboard blanket were discarded. The key point was how to find the engineering design solution of the TF coil system with the high field and high current density. The coil system with the center post radius of less than 1 m can generate the maximum field of ∼ 20 T. This coil system causes a compact reactor concept, where the plasma major and minor radii of 3.75 m and 1.9 m, respectively and the fusion power of 1.8 GW. (author)
Tewes, Philipp; Genschow, Konstantin; Little, Jesse; Wygnanski, Israel
2017-11-01
A detailed flow survey using PIV was conducted over a highly-deflected flap (55°) of a low-aspect ratio trapezoidal wing. The wing section is a NACA 0012 with 45° sweep at both the leading and trailing edges, an aspect ratio of 1.5 and a taper ratio of 0.27. The main element is equipped with 7 equally spaced fluidic oscillators, covering the inner 60 % of the span, located near the flap hinge. Experiments were carried out at 0° and 8° incidence at a Reynolds number of 1.7 .106 for both baseline and active flow control (AFC) cases. Velocity ISO-surfaces, x-vorticity and streamlines are analyzed / discussed. A flap leading edge vortex governs the baseline flow field for 0°. This vortical structure interacts with the jets emitted by the actuators (Cμ = 1 %). Its development is hampered and the vortex is redirected toward the trailing edge resulting in a CL increase. At 8°, the dominant flap leading edge vortex could not be detected and is believed to have already merged with the tip vortex. AFC attached the flow over the flap and enhanced the lift by up to 20 % while maintaining longitudinal stability. The dominant flow features in the AFC cases are actuator-generated streamwise vortices which appear stronger at 8°. This work was supported by the Office of Naval Research under ONR Grant No. N00014-14-1-0387.
Engineering feasibility of tight aspect ratio Tokamak (spherical torus) reactors
International Nuclear Information System (INIS)
Peng, Y-K.M.; Hicks, J.B.
1990-01-01
Engineering solutions are identified and analyzed for key high-power-density components of tight aspect ratio tokamak reactors (spherical torus reactors). The potentially extreme divertor heat loads can be reduced to about 3 MW/m 2 in expanded divertors using coils inside the demountable toroidal field coils. Given the long and narrow divertor channels, gaseous divertor targets become possible, which eliminate sputtering and increase the divertor life. The unshielded centre conductor post (CCP) of the toroidal field coil can be made of a single dispersion strengthened copper conductor cooled by high-velocity pressurized water to maintain acceptable copper temperature and strength. Damage and activation of the CCP at a neutron fluence of 10 MW-a/m 2 are also tolerable. Annual replacement of the centre post, the divertor assemblies and the blanket can be accomplished with vertical access for all torus components, which are modularized to reduce size and weight. The technical requirements of these solutions are shown to be comparable with, if not less demanding than, those estimated for conventional tokamak reactors. (author)
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N. Sitaram
2011-01-01
Full Text Available The flow field at the rotor exit of a low aspect ratio axial flow fan for different tip geometries and for different flow coefficients is measured in the present study. The following configurations are tested: (1 rotor without partial shroud, designated as rotor (wos, (2 rotor with partial shroud, designated as rotor (ws, and (3 rotor with perforated (perforations in the shape of discrete circular holes partial shroud, designated as rotor (wps. From steady state measurements, the performance of rotor (wps is found to be the best. Both the rotors with partial shrouds have stalled at a higher flow coefficient compared to that of rotor (wos. From periodic flow measurements, it is concluded that the low velocity region near the tip section is considerably reduced with the use of partial shrouds with perforations. The extent of this low velocity region for both rotor (wos and rotor (wps increases with decreasing flow coefficient due to increased stage loading. This core of low momentum fluid has moved inwards of the annulus and towards the pressure side as the flow coefficient decreases. The extent of the low momentum fluid is smaller for rotor (wps than that of rotor (wos at all flow coefficients.
Tokamak reactor designs as a function of aspect ratio
International Nuclear Information System (INIS)
Wong, C.P.C.; Stambaugh, R.D.
2000-01-01
This paper assesses the technical and economic potential of tokamak power plants which utilize superconducting coil (SC) or normal conducting coil (NC) designs as a function of aspect ratio (A). Based on the results from plasma equilibrium calculations, the key physics design parameters of β N , β p , β T , and κ were fitted to parametric equations covering A in the range of 1.2-6. By using ARIES-RS and ARIES-ST as reference design points, a fusion reactor system code was used to project the performance and cost of electricity (COE) of SC and NC reactor designs over the same range of A. The principle difference between the SC and the NC designs are the inboard standoff distance between the coil and the inboard first wall, and the maximum central column current density used for respective coil types. Results show that at an output power of 2 GWe both NC and SC designs can project COE in the respectable range of 62-65 mill/kW h at gross thermal efficiency of 46%, with neutron wall loading (Γ n ) ∼7 MW/m 2 . More importantly, we have learned that based on the present knowledge of equilibrium physics and fusion power core components and system design we can project the performance and COE of reactor designs at least for the purpose of comparative assessment. Tokamak design points can then be selected and optimized for testing or commercial devices as a function of output power, A and Γ n for both SC and NC design options
Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD
Energy Technology Data Exchange (ETDEWEB)
Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of); Chung, Kyoo Sun [Hanyang Univ., Seoul (Korea, Republic of); Hong, Sang Heui [Seoul National Univ., Seoul (Korea, Republic of); Kang, Heui Dong [Kyungpook National Univ., Taegu (Korea, Republic of); Lee, Jae Koo [Pohang Inst. of Science and Technology, Kyungnam (Korea, Republic of)
1994-11-01
A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the `advanced tokamak` physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs.
Concept definition of KT-2, a large-aspect-ratio diverter tokamak with FWCD
International Nuclear Information System (INIS)
Kim, Sung Kyoo; Chang, In Soon; Chung, Moon Kyoo; Hwang, Chul Kyoo; Lee, Kwang Won; In, Sang Ryul; Choi, Byung Ho; Hong, Bong Keun; Oh, Byung Hoon; Chung, Seung Ho; Yoon, Byung Joo; Yoon, Jae Sung; Song, Woo Sub; Chang, Choong Suk; Chang, Hong Yung; Choi, Duk In; Nam, Chang Heui; Chung, Kyoo Sun; Hong, Sang Heui; Kang, Heui Dong; Lee, Jae Koo
1994-11-01
A concept definition of the KT-2 tokamak is made. The research goal of the machine is to study the 'advanced tokamak' physics and engineering issues on the mid size large-aspect-ratio diverter tokamak with intense RF heating (>5 MW). Survey of the status of the research fields, the physics basis for the concept, operation scenarios, as well as machine design concept are presented. (Author) 86 refs., 17 figs., 22 tabs
International Nuclear Information System (INIS)
Wesson, John.
1996-01-01
This book is the first compiled collection about tokamak. At first chapter tokamak is represented from fusion point of view and also the necessary conditions for producing power. The following chapters are represent plasma physics, the specifications of tokamak, plasma heating procedures and problems related to it, equilibrium, confinement, magnetohydrodynamic stability, instabilities, plasma material interaction, plasma measurement and experiments regarding to tokamak; an addendum is also given at the end of the book
International Nuclear Information System (INIS)
Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.
1997-01-01
The possible use of tokamaks for thermonuclear power plants is discussed, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First, the existing literature is reviewed and summarized. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamaks power plant, by including the power required to drive the toroidal field and by considering two extremes of plasma current drive efficiency. Third, the analytic results are augmented by a numerical calculation that permits arbitrary plasma current drive efficiency and different confinement scaling relationships. Throughout, the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculation of electric power. The latest published reactor studies show little advantage in using low aspect ratios to obtain a more compact device (and a low cost of electricity) unless either remarkably high efficiency plasma current drive and low safety factor are combined, or unless confinement (the H factor), the permissible elongation and the permissible neutron wall loading increase as the aspect ratio is reduced. These results are reproduced with the analytic model. (author). 22 refs, 3 figs
International Nuclear Information System (INIS)
Jarboe, T.R.; Nelson, B.A.
1997-01-01
During the reporting period of this HIT grant (1 June 1995--31 May 1997) we've conducted further stability analysis, used the TIP diagnostic to measure plasma fields in HIT, and developed a single-parameter helicity injector model. HIT has undergone a significant upgrade to the HIT-II configuration which is described here. Parts for HIT-II have been designed, ordered, and received under this grant and are being assembled under the subsequent grant
Destruction of magnetic surfaces in the edge of a large aspect ratio Tokamak with ergodic limiter
International Nuclear Information System (INIS)
Viana, R.L.; Caldas, I.L.
1990-01-01
The model of Martin and Taylor for a large aspect-ratio Tokamak with an ergodic limiter is considered. In order to study the onset of chaotic behaviour for the magnetic field lines in the edge of the vessel, a Hamiltonian formulation is constructed for the system and the overlap of two peripheral magnetic islands is considered. So, it is possible to determine a threshold for the ergodic limiter current to cause destruction of rational magnetic surfaces in this region. (Author)
International Nuclear Information System (INIS)
He Qibing; Peng Qiyang; Qu Wenxiao
1993-09-01
The effect of energetic trapped particles on the stabilization of ballooning modes in finite-aspect-ratio tokamaks is numerically analyzed. The numerical solution of boundary value problem of an integro-differential equation is successfully obtained by RKF integral method with variable step size. The results show that the instability domain of ballooning modes becomes small along with the increase of energetic particles pressure. The energetic trapped particles can partially or completely suppress the instability of ballooning modes
Development of Operation Scenario for Spherical Tokamak at SNU
International Nuclear Information System (INIS)
Sung, C. K.; Park, Y. S.; Lee, H. Y.; Kang, J.; Hwang, Y. S.
2009-01-01
Several concepts for nuclear fusion plant exist. In these concepts, tokamak is the most promising one to realize nuclear fusion plant. Though tokamak has leading concept, and this has world record in fusion heating power, tokamak has the critical drawback: low heating efficiency. That is the reason why we need another alternative concept which compensates tokamak's disadvantage. Spherical Torus(ST) is one of these kinds of concepts. ST is a kind of tokamak which has low aspect ratio. This feature gives ST advantages compared to conventional tokamak: high efficiency, compactness, low cost. However, ST lacks central region for solenoid that is needed to start-up and sustain. Since it is the most efficient that initializing and sustaining by using solenoid, this is ST's intrinsic limitation. To overcome this, a new device which can start-up and sustain ST plasmas by means of continuous tokamak plasma injection has been designed
Diagnostics systems for the TBR-E tokamak
International Nuclear Information System (INIS)
Ueda, M.; Ferreira, J.L.; Aso, Y.; Ferreira, J.G.
1992-08-01
A general view of the several diagnostics systems proposed for the TBR-E tokamak is given. This project is a joint undertaking of INPE, USP and UNICAMP plasma laboratories. The requirements for the measurements of the plasma produced parameters are described. Special attention is given for diagnostics used to investigate new physical issues on a low aspect ratio tokamak such as TBR-E. (author)
Comments on the asymptotic treatment of tokamak MHD-stability at large aspect ratio
International Nuclear Information System (INIS)
Rebhan, E.
1980-01-01
In the asymptotic treatment of tokamak MHD stability at small inverse aspect ratio epsilon, the special case of poloidal wave number m=0 has been treated improperly in the literature for both axisymmetric and non-axisymmetric modes. In axisymmetric stability, a contribution to the perturbational vacuum field is either omitted or cancelled. In a variational stability analysis this field contribution provides σ 2 W with a correction term proportional to (1nepsilon) -1 , which may change the asymptotic range of stability and improve agreement with numerical finite-aspect-ratio results. In non-axisymmetric stability, for the perturbational vacuum field of the m=0 modes, usually the wrong of two possible solutions is chosen. It is shown why in many cases this wrong choice has no consequences on the correctness of the stability results, and circumstances are pointed out under which consequences may arise. (author)
Basic toroidal Effects on Alfven Wave Current in Small Aspect Ratio Tokamaks
International Nuclear Information System (INIS)
Burma, C.; Cuperman, S.; Komoshvili, K.
1998-01-01
The Alfven wave current drive (AWCD) in small aspect ratio Tokamaks is properly calculated, with consideration of the basic toroidicity effects present in (i) the dielectric tensor-operator (involving the strongly toroidal equilibrium profiles), (ii) the structure of the r.f. fields obtained as a solution of the wave equation (through Maxwell's equations' toroidal operators as well as the conversion rate and conversion layer location, depending also on the equilibrium profiles) and (iii) the formulation of the AWCD (which, besides its dependence on the r.f. fields - affected by toroidicity as mentioned at points (i) and (ii) - also requires the equilibrium-magnetic-surface averaging of non-resonant forces involved). Thus, we consider consistent equilibrium profiles with neo-classical conductivity corresponding to an ohmic START-like discharge; use a resistive (anisotropic) MHD dielectric tensor-operator Edith practically no limitations, adequate to describe the plasma response in the pre-heated stage ; solve numerically the 2(1/2)D full- wave equation by the aid of an advanced finite element code developed in; and evaluate the AWCD by the aid of the recently proposed, quite general formulation holding in the case of strongly toroidal fusion devices and including contributions due to helicity injection, momentum transfer and plasma Bow. A general discussion of the results obtained in this work is presented
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1999-01-01
Some basic aspects of wave-plasma interaction of interest for tight aspect ratio spherical tokamaks are investigated theoretically. The following scenario is considered: A. Fast magnetosonic waves are launched by an external antenna into a simulated spherical Tokamak plasma; these waves are converted to Alfven waves at points (layer) satisfying the Alfven resonance condition. B. The simulated spherical tokamaks-plasma has a circular cross-section and toroidicity effects are simulated by Grad-Shafranov type, radially dependent axial magnetic field and its shear. (J. Actual equilibrium profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. D. The study is based on the numerical solution of the full e.m. wave equation which includes a quite general resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. Two kinds of results will be presented: I. Proofs validating the computational algorithm used and including convergence and energy conservation. II. Exact quantitative results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited p over . The dependence of the results on the launched wave characteristics (wave numbers, frequency and intensity) as well as on those of the equilibrium plasma (equilibrium current, neoclassical resistivity and electron inertia) will be discussed
Compact tokamak reactors. Part 1 (analytic results)
International Nuclear Information System (INIS)
Wootton, A.J.; Wiley, J.C.; Edmonds, P.H.; Ross, D.W.
1996-01-01
We discuss the possible use of tokamaks for thermonuclear power plants, in particular tokamaks with low aspect ratio and copper toroidal field coils. Three approaches are presented. First we review and summarize the existing literature. Second, using simple analytic estimates, the size of the smallest tokamak to produce an ignited plasma is derived. This steady state energy balance analysis is then extended to determine the smallest tokamak power plant, by including the power required to drive the toroidal field, and considering two extremes of plasma current drive efficiency. The analytic results will be augmented by a numerical calculation which permits arbitrary plasma current drive efficiency; the results of which will be presented in Part II. Third, a scaling from any given reference reactor design to a copper toroidal field coil device is discussed. Throughout the paper the importance of various restrictions is emphasized, in particular plasma current drive efficiency, plasma confinement, plasma safety factor, plasma elongation, plasma beta, neutron wall loading, blanket availability and recirculating electric power. We conclude that the latest published reactor studies, which show little advantage in using low aspect ratio unless remarkably high efficiency plasma current drive and low safety factor are combined, can be reproduced with the analytic model
Effects of magnetic shear on current penetration in a tokamak
International Nuclear Information System (INIS)
Zhang Pengyun; Wang Long
2001-01-01
The penetrations of the parallel and perpendicular components of plasma currents are interrelated to each other due to the existence of magnetic shear in a tokamak configuration. Effects of the shear on the penetration of Fourier components of toroidal plasma current are analysed in a cylindrical column model. The current penetration is obviously strengthened by the shear for a bell-bike conductivity profile and low safety factor and low aspect ratio
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
1999-01-01
Theoretical results on the wave-plasma interactions in simulated toroidal configurations are presented. The study covers the cases of large to low aspect ratio tokamaks, in the pre-heated stage. Fast waves emitted from an external antenna with different wave numbers and frequencies are considered. The non-inductive Alfven wave current drive is evaluated and discussed. (author)
International Nuclear Information System (INIS)
Cuperman, S.; Bruma, C.; Komoshvili, K.
2001-01-01
Theoretical results on the wave-plasma interactions in simulated toroidal configurations are presented. The study covers the cases of large to low aspect ratio tokamaks, in the pre-heated stage. Fast waves emitted from an external antenna with different wave numbers and frequencies are considered. The non-inductive Alfven wave current drive is evaluated and discussed. (author)
International Nuclear Information System (INIS)
Bruma, C.; Cuperman, S.; Komoshvili, K.
1996-01-01
Supplementary non-inductive current drive and heating are necessary to bring Tokamak plasmas into the ignition regime. The resonant excitation of shear Alfven waves (SAW) - in the continuum range (CR) or/and in the discrete global Alfven eigenmode spectrum (GAE's) - represents one potential, suitable method for this purpose. Within the framework of ideal MHD, the current drive (CD) via helicity injection in Tokamak plasmas has been considered by Cuperman et al (1996) and Komoshvili et al. (1996). This work is concerned with the investigation of the non-ideal resistive MHD effects on both the excitation of SAW's (CR's and GAE's) and the generation of non-inductive current drive via helicity injection in Tokamak plasmas. The research covers Tokamak aspect ratios ranging between large value cases (R/a = 10) and the very tight value case (R/ a = 1.2). (authors)
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Komoshvili, K. [Tel Aviv Univ. (Israel). School of Physics and Astronomy; Coll. of Judea and Samaria, Ariel (Israel); Cuperman, S. [Tel Aviv Univ. (Israel). School of Physics and Astronomy
2000-11-01
Some basic aspects of wave-plasma interaction of special interest for tight aspect ratio (spherical) tokamaks (ST's) are investigated numerically; these aspects include fast mode conversion and energy deposition. The study is based on the numerical solution of the full electro-magnetic (e.m.) wave equation which includes a quite general two-fluid, resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. A generalized expression for the power absorption appropriate for the above scenario, with consideration of all the basic effects also present in the dielectric tensor-operator, was derived and used. The current-carrying ST-plasma has a circular cross-section and toroidicity effects are simulated by a Grad-Shafranov type, radially dependent axial magnetic field and its shear; however, the Shafranov shift is not considered. Actually, the equilibrium parameters and radial profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. Fast magnetosonic waves are launched from an external antenna into this simulated spherical tokamak plasma; these waves are converted to Alfven waves at points (layers) satisfying the Alfven resonance condition. Quantitative-results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited power are presented. Their dependence on the equilibrium plasma current, neoclassical resistivity and electron inertia as well as on those of the antenna launched wave (wave numbers, frequency and current intensity) is systematically studied and discussed. (orig.)
International Nuclear Information System (INIS)
Bruma, C.; Komoshvili, K.; Cuperman, S.
2000-01-01
Some basic aspects of wave-plasma interaction of special interest for tight aspect ratio (spherical) tokamaks (ST's) are investigated numerically; these aspects include fast mode conversion and energy deposition. The study is based on the numerical solution of the full electro-magnetic (e.m.) wave equation which includes a quite general two-fluid, resistive MHD dielectric tensor, with consideration of equilibrium current and neoclassical effects. A generalized expression for the power absorption appropriate for the above scenario, with consideration of all the basic effects also present in the dielectric tensor-operator, was derived and used. The current-carrying ST-plasma has a circular cross-section and toroidicity effects are simulated by a Grad-Shafranov type, radially dependent axial magnetic field and its shear; however, the Shafranov shift is not considered. Actually, the equilibrium parameters and radial profiles (magnetic field, pressure and current) observed in the low field side (LFS) of spherical tokamaks (viz., START at Culham, UK) are used. Fast magnetosonic waves are launched from an external antenna into this simulated spherical tokamak plasma; these waves are converted to Alfven waves at points (layers) satisfying the Alfven resonance condition. Quantitative-results concerning (i) the structure and space dependence of the mode-converted Alfven waves and (ii) the basic features of the deposited power are presented. Their dependence on the equilibrium plasma current, neoclassical resistivity and electron inertia as well as on those of the antenna launched wave (wave numbers, frequency and current intensity) is systematically studied and discussed. (orig.)
Power plant design study of a high aspect ratio Tokamak using a SiC composite structure
International Nuclear Information System (INIS)
Murakami, Y.; Takase, H.; Shinya, K.
1998-01-01
The DREAM (drastically easy maintenance) tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio (A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high-A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m -2 , the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m -2 . A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition. (orig.)
International Nuclear Information System (INIS)
Peng, Y.K.M.
1978-04-01
A tokamak configuration is proposed that permits the rapid replacement of a plasma discharge in a ''burn'' chamber by another one in a time scale much shorter than the elementary thermal time constant of the chamber first wall. With respect to the chamber, the effective duty cycle factor can thus be made arbitrarily close to unity minimizing the cyclic thermal stress in the first wall. At least one plasma discharge always exists in the new tokamak configuration, hence, a continuous tokamak. By incorporating adiabatic toroidal compression, configurations of continuous tokamak compressors are introduced. To operate continuous tokamaks, it is necessary to introduce the concept of mixed poloidal field coils, which spatially groups all the poloidal field coils into three sets, all contributing simultaneously to inducing the plasma current and maintaining the proper plasma shape and position. Preliminary numerical calculations of axisymmetric MHD equilibria in continuous tokamaks indicate the feasibility of their continued plasma operation. Advanced concepts of continuous tokamaks to reduce the topological complexity and to allow the burn plasma aspect ratio to decrease for increased beta are then suggested
Unsteady transonic flow analysis for low aspect ratio, pointed wings.
Kimble, K. R.; Ruo, S. Y.; Wu, J. M.; Liu, D. Y.
1973-01-01
Oswatitsch and Keune's parabolic method for steady transonic flow is applied and extended to thin slender wings oscillating in the sonic flow field. The parabolic constant for the wing was determined from the equivalent body of revolution. Laplace transform methods were used to derive the asymptotic equations for pressure coefficient, and the Adams-Sears iterative procedure was employed to solve the equations. A computer program was developed to find the pressure distributions, generalized force coefficients, and stability derivatives for delta, convex, and concave wing planforms.
International Nuclear Information System (INIS)
Burma, C.; Cuperman, S.; Komoshvili, K.
1998-01-01
The wave equation for strongly toroidal small aspect ratio (spherical) tokamaks with non-circular cross-section is properly formulated and solved for global waves, in the Alfven frequency range. The current-carrying toroidal plasma is surrounded by a helical sheet-current antenna, which is enclosed within a perfectly conducting wall. The problem is formulated in terms of the vector and scalar potentials (A,Φ), thus avoiding the numerical solution occurring in the case of (E,B) formulation. Adequate boundary conditions are applied at the vacuum - metallic wall interface and the magnetic axis. A recently derived dielectric tensor-operator, able to describe the anisotropic plasma response in spherical tokamaks, is used for this purpose; except for its linear character, no physical or geometrical limitations are imposed on it. The equilibrium profiles (magnetic field, pressure and current) are obtained from a numerical solution of the Grad-Shafranov equation. Specifically, the wave equation is solved by the aid of a numerical code we developed for the present problem, based on the well documented 2(1/2)D finite element solver proposed by E.G. Sewell. With the definitions V i (θ,ρ) = U i (-θ,ρ) (V i U i = A j , Φ; j = ρ,φ,θ), our code solves simultaneously 16 second order partial differential equations (eight equations for each of real and imaginary set of functions V i , U i ). A systematic analysis of the solutions obtained for various values and combinations of wavenumbers and frequencies in the Alfven range is presented
Stability of highly shifted equilibria in a large aspect ratio low-field tokamak
International Nuclear Information System (INIS)
Gourdain, P.-A.; Leboeuf, J.-N.; Neches, R. Y.
2007-01-01
In the long run, the economics of fusion will dictate that reactors confine large plasma pressure rather efficiently. A possible route manifests itself as equilibria with large shift of the plasma magnetic axis. This shift compresses the flux surfaces on the outer part of the plasma, hereby increasing the allowable plasma pressure a machine can confine for a given toroidal magnetic field, which is the main cost of the device. As a first step toward a reactor, we propose investigating the stability of such configurations in a low magnetic field high aspect ratio machine. By focusing our arguments solely on the shape of the toroidal plasma current density profile we discuss the stability of highly shifted equilibria and their robustness to current profile variations that could occur in actual experiments. The evolution of the plasma parameters, as the beta poloidal is increased, is also examined to give a better understanding of the difference in performance between the various regimes
Tokamak transmutation of (nuclear) waste (TTW): Parametric studies
International Nuclear Information System (INIS)
Cheng, E.T.; Krakowski, R.A.; Peng, Y.K.M.
1994-01-01
Radioactive waste generated as part of the commercial-power and defense nuclear programs can be either stored or transmuted. The latter treatment requires a capital-intensive neutron source and is reserved for particularly hazardous and long-lived actinide and fission-product waste. A comparative description of fusion-based transmutation is made on the basis of rudimentary estimates of ergonic performance and transmutation capacities versus inventories for both ultra-low-aspect-ratio (spherical torus, ST) and conversional (aspect-ratio) tokamak fusion-power-core drivers. The parametric systems studies reported herein provides a preamble to more-detailed, cost-based systems analyses
Baber, Hal T , Jr; Moul, Martin T
1955-01-01
Two full-scale models of an inline, cruciform, canard missile configuration having a low-aspect-ratio wing equipped with flap-type controls were flight tested in order to determine the missile's longitudinal aerodynamic characteristics. Stability derivatives and control and drag characteristics are presented for a range of Mach number from 0.7 to 1.8. Nonlinear lift and moment curves were noted for the angle - of-attack range of this test (0 deg to 8 deg). The aerodynamic-center location for angles of attack near 50 remained nearly constant for supersonic speeds at 13.5 percent of the mean aerodynamic chord; whereas for angles of attack near 0 deg, there was a rapid forward movement of the aerodynamic center as the Mach number increased. At a control deflection of 0 deg, the missile's response to the longitudinal control was in an essentially fixed space plane which was not coincident with the pitch plane as a result of the missile rolling. As a consequence, stability characteristics were determined from the resultant of pitch and yaw motions. The damping-in-pitch derivatives for the two angle -of-attack ranges of the test are in close agreement and varied only slightly with Mach number. The horn-balanced trailing-edge flap was effective in producing angle of attack over the Mach number range.
Brown, Clarence A , Jr
1957-01-01
A full- scale rocket-powered model of a cruciform canard missile configuration with a low- aspect - ratio wing and blunt nose has been flight tested by the Langley Pilotless Aircraft Research Division. Static and dynamic longitudinal stability and control derivatives of this interdigitated canard-wing missile configuration were determined by using the pulsed- control technique at low angles of attack and for a Mach number range of 1.2 to 2.1. The lift - curve slope showed only small nonlinearities with changes in control deflection or angle of attack but indicated a difference in lift- .curve slope of approximately 7 percent for the two control deflections of delta = 3.0 deg and delta= -0.3 deg . The large tail length of the missile tested was effective in producing damping in pitch throughout the Mach number range tested. The aerodynamic- center location was nearly constant with Mach number for the two control deflections but was shown to be less stable with the larger control deflection. The increment of lift produced by the controls was small and positive throughout the Mach number range tested, whereas the pitching moment produced by the controls exhibited a normal trend of reduced effectiveness with increasing Mach number.The effectiveness of the controls in producing angle of attack, lift, and pitching moment was good at all Mach numbers tested.
Baber, H. T., Jr.; Moul, M. T.
1955-01-01
Two full-scale models of an inline, cruciform, canard missile configuration having a low-aspect-ratio wing equipped with flap-type controls were flight tested in order to determine the missile's longitudinal aerodynamic characteristics. Stability derivatives and control and drag characteristics are presented for a range of Mach number from 0.7 to 1.8. Nonlinear lift and moment curves were noted for the angle-of-attack range of this test (0 deg to 8 deg ). The aerodynamic-center location for angles of attack near 5 deg remained nearly constant for supersonic speeds at 13.5 percent of the mean aerodynamic chord; whereas for angles of attack near O deg, there was a rapid forward movement of the aerodynamic center as the Mach number increased. At a control deflection of O deg, the missile's response to the longitudinal control was in an essentially fixed space plane which was not coincident with the pitch plane as a result of the missile rolling. As a consequence, stability characteristics were determined from the resultant of pitch and yaw motions. The damping-in-pitch derivatives for the two angle-of-attack ranges of the test are in close agreement and varied only slightly with Mach number. The horn-balanced trailing-edge flap was effective in producing angle of attack over the Mach number range.
The spheric tokamak programme at Culham
International Nuclear Information System (INIS)
Sykes, A.
1999-01-01
The Spherical Tokamak (ST) is the low aspect ratio limit of the conventional tokamak, and appears to offer attractive physics properties in a simpler device. The START (Small Tight Aspect Ratio Tokamak) experiment provided the world's first demonstration of the properties of hot plasmas in an ST configuration, and was operational at Culham from January 1991 to March 1998, obtaining plasma current of up to 300 kA and pulse durations of ∼ 50 ms. Its successor, MAST is scheduled to obtain first plasma in Autumn 1998 and is a purpose built, high vacuum machine designed to have a tenfold increase in plasma volume with plasma currents up to 2 MA. Current drive and heating will be by a combination of induction-compression as on START, a high-performance central solenoid, 1.5 MW ECRH and 5 MW of Neutral Beam Injection. The promising results from START are reviewed, and the many challenges posed for the next generation of purpose-built STs (such as MAST) are described. (author)
First physics results from the MAST Mega-Amp Spherical Tokamak
International Nuclear Information System (INIS)
Sykes, A.; Ahn, J.-W.; Akers, R.; Arends, E.; Carolan, P.G.; Counsell, G.F.; Fielding, S. J.; Gryaznevich, M.; Martin, R.; Price, M.; Roach, C.; Shevchenko, V.; Tournianski, M.; Valovic, M.; Walsh, M.J.; Wilson, H.R.
2001-01-01
First physics results are presented from MAST (Mega-Amp Spherical Tokamak), one of the new generation of purpose built spherical tokamaks (STs) now commencing operation. Some of these results demonstrate, for the first time, the novel effects of low aspect ratio, for example, the enhancement of resistivity due to neo-classical effects. H-mode is achieved and the transition to H-mode is accompanied by a tenfold steepening of the edge density gradient which may enable the successful application of electron Bernstein wave heating in STs. Studies of halo currents show that these less than expected from conventional tokamak results, and measurements of divertor power loading confirm that most of the power flows to the outer strike points, easing the power handling on the inner points (a critical issue for STs)
Simulation of saturated tearing modes in tokamaks
International Nuclear Information System (INIS)
Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.
2004-01-01
A quasi-linear model, which includes the effect of the neoclassical bootstrap current, is developed for saturated tearing modes in order to compute magnetic island widths in axisymmetric toroidal plasmas with arbitrary aspect ratio and cross-sectional shape. The model is tested in a simple stand-alone code and is implemented in the BALDUR [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1982)] predictive modeling code. It is found that the widths of tearing mode islands increase with decreasing aspect ratio and with increasing elongation. Also, the island widths increase when the gradient of the current density increases at the edge of the islands and when the current density inside the islands is suppressed, such as the suppression caused by the near absence of the bootstrap current within the islands. In simulations of tokamak discharges, it is found that tearing mode island widths oscillate in time in response to periodic sawtooth crashes. The local enhancements in the transport produced by magnetic islands have a noticeable effect on global plasma confinement in simulations of low aspect ratio, high beta tokamaks, where saturated tearing mode islands can occur with widths that are greater than 15% of the plasma minor radius
Optimization study of normal conductor tokamak for commercial neutron source
Fujita, T.; Sakai, R.; Okamoto, A.
2017-05-01
The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.
National Aeronautics and Space Administration — The performance gains and weight reductions from using Ceramic Matrix Composite(CMC) turbine blades in both the High Pressure Turbine(HPT) and Low Pressure...
The Ar{sup 17+} Ly{sub {alpha}2}/Ly{sub {alpha}1} ratio in Alcator C-Mod tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Rice, J E; Reinke, M L; Ince-Cushman, A C; Podpaly, Y A [Plasma Science and Fusion Center, MIT, Cambridge, MA (United States); Ashbourn, J M A [Mathematical Institute, University of Oxford, Oxford (United Kingdom); Gu, M F [SSL, University of California Berkeley, CA (United States); Bitter, M; Hill, K [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Rachlew, E, E-mail: rice@psfc.mit.edu [KTH, Stockholm (Sweden)
2011-08-28
High-quality spectra of hydrogen-like Ar{sup 17+} have been obtained from Alcator C-Mod tokamak plasmas using a spatially imaging high-resolution x-ray spectrometer system in an extensive study of the underlying high-n satellite lines. The ratio of Ly{sub {alpha}2} (1S{sub 1/2}-2P{sub 1/2}) to Ly{sub {alpha}1} (1S{sub 1/2}-2P{sub 3/2}) was found to be {approx}0.52 regardless of plasma parameters, which is somewhat greater than the ratio of the statistical weights of the upper n = 2 levels, 0.5. This difference is mainly due to the effects of collisional excitation of fine-structure sub-levels. For the observations presented here, electron densities were in an extended range from 3x10{sup 19} to 4x10{sup 20} m{sup -3} with electron and ion temperatures between 1 and 4 keV. Experimental results are compared to calculations from COLRAD, a collisional-radiative modelling code, and good agreement is shown.
Interaction of a spheromak-like compact toroid with a high beta spherical tokamak plasma
International Nuclear Information System (INIS)
Hwang, D.Q.; McLean, H.S.; Baker, K.L.; Evans, R.W.; Horton, R.D.; Terry, S.D.; Howard, S.; Schmidt, G.L.
2000-01-01
Recent experiments using accelerated spheromak-like compact toroids (SCTs) to fuel tokamak plasmas have quantified the penetration mechanism in the low beta regime; i.e. external magnetic field pressure dominates plasma thermal pressure. However, fusion reactor designs require high beta plasma and, more importantly, the proper plasma pressure profile. Here, the effect of the plasma pressure profile on SCT penetration, specifically, the effect of diamagnetism, is addressed. It is estimated that magnetic field pressure dominates penetration even up to 50% local beta. The combination of the diamagnetic effect on the toroidal magnetic field and the strong poloidal field at the outer major radius of a spherical tokamak will result in a diamagnetic well in the total magnetic field. Therefore, the spherical tokamak is a good candidate to test the potential trapping of an SCT in a high beta diamagnetic well. The diamagnetic effects of a high beta spherical tokamak discharge (low aspect ratio) are computed. To test the penetration of an SCT into such a diamagnetic well, experiments have been conducted of SCT injection into a vacuum field structure which simulates the diamagnetic field effect of a high beta tokamak. The diamagnetic field gradient length is substantially shorter than that of the toroidal field of the tokamak, and the results show that it can still improve the penetration of the SCT. Finally, analytic results have been used to estimate the effect of plasma pressure on penetration, and the effect of plasma pressure was found to be small in comparison with the magnetic field pressure. The penetration condition for a vacuum field only is reported. To study the diamagnetic effect in a high beta plasma, additional experiments need to be carried out on a high beta spherical tokamak. (author)
START: the creation of a spherical tokamak
International Nuclear Information System (INIS)
Sykes, Alan
1992-01-01
The START (Small Tight Aspect Ratio Tokamak) plasma fusion experiment is now operational at AEA Fusion's Culham Laboratory. It is the world's first experiment to explore an extreme limit of the tokamak - the Spherical Tokamak - which theoretical studies predict may have substantial advantages in the search for economic fusion power. The Head of the START project, describes the concept, some of the initial experimental results and the possibility of developing a spherical tokamak power reactor. (author)
International Nuclear Information System (INIS)
Berger, D.; Bernard, L.C.; Gruber, R.; Troyon, F.
1980-01-01
The Lausanne ideal MHD stability code ERATO is used to investigate spectral properties of Solovev's equilibrium at small aspect ratios. Two different elongations are considered. Both free and rigid boundary models are computed and compared. Modes characterized by a large radial extension have been found which appear to be due to coupling of m=1 and m=2 modes due to toricity. The internal modes spectrum is compared with the predictions of the full Mercier criterion, taking into account its spatial dependence, and with the ballooning modes stability criterion. (Auth.)
First Results from Tests of High Temperature Superconductor Magnets on Tokamak
Energy Technology Data Exchange (ETDEWEB)
Gryaznevich, M.; Todd, T.T., E-mail: mikhail.gryaznevich@ccfe.ac.uk [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Svoboda, V.; Markovic, T.; Ondrej, G. [Czech Technical University, Prague (Czech Republic); Stockel, J.; Duran, I.; Kovarik, K. [IPP Prague, Czech Technical University, Prague (Czech Republic); Sykes, A.; Kingham, D. [Tokamak Solutions, Culham Science Centre, Abingdon (United Kingdom); Melhem, Z.; Ball, S.; Chappell, S. [Oxford Instruments, Abingdon (United Kingdom); Lilley, M. K.; De Grouchy, P.; Kim, H. -T. [Imperial College, London (United Kingdom)
2012-09-15
construction of a small fully-HTS low aspect ratio tokamak has started at the Tokamak Solutions UK premises in the Culham Science Centre. It is planned to operate a small tokamak with A = 2 and circular cross section in steady state with plasma currents of 10 - 20 kA driven by Electron Bernstein Wave current drive. In parallel, the design and manufacture of a high-field (5 T) HTS TF coil for a spherical tokamak is carried out. (author)
New directions in tokamak reactors
International Nuclear Information System (INIS)
Baker, C.C.
1985-01-01
New directions for tokamak research are briefly mentioned. Some of the areas for new considerations are the following: reactor size, beta ratio, current drivers, blankets, impurity control, and modular designs
International Nuclear Information System (INIS)
Robinson, D.C.
1987-01-01
With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt
Energy Technology Data Exchange (ETDEWEB)
Bruma, C.; Cuperman, S.; Komoshvili, K. [Tel Aviv Univ., Ramat Aviv (Israel)
2005-08-01
As it is the case with tokamaks in general, and moreover, due to their specific geometry (limited space for inboard solenoid magnets), low aspect ratio (spherical) tokamaks (STs) require additional auxiliary non-ohmic current startup and maintenance, generation of internal transport barriers (associated with underlying sheared poloidal flows and quasi-stationary radial electric fields), plasma heating, etc. One of the options to generate these necessary effects in STs is by the aid of rf waves launched from a suitable external antenna; in this option the effects just mentioned are a consequence of ponderomotive forces resulting from the interaction of the rf waves with the plasma. Since experimental data on STs (viz., the START-device) reveal the presence of an anomalous plasma resistivity (about four times Spitzer's one), we carried out a systematic parametric investigation of the effects of an increased plasma resistivity on the magnitude and spatial localization of the resulting power deposition.
Unified Ideal Stability Limits for Advanced Tokamak and Spherical Torus Plasmas
International Nuclear Information System (INIS)
Menard, J.E.; Bell, M.G.; Bell, R.E.; Gates, D.A.; Kaye, S.M.; LeBlanc, B.P.; Sabbagh, S.A.; Fredrickson, E.D.; Jardin, S.C.; Maingi, R.; Manickam, J.; Mueller, D.; Ono, M.; Paoletti, F.; Peng, Y.-K.M.; Soukhanovskii, V.; Stutman, D.; Synakowski, E.J.
2003-01-01
Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal nonrotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants
Kirschmeier, Benjamin; Summerour, Jacob; Bryant, Matthew
2017-04-01
Interest in clean, stable, and renewable energy harvesting devices has increased dramatically with the volatility of petroleum markets. Specifically, research in aero/hydro kinetic devices has created numerous new horizontal and vertical axis wind turbines, and oscillating wing turbines. Oscillating wing turbines (OWTs) differ from their wind turbine cousins by having a rectangular swept area compared to a circular swept area. The OWT systems also possess a lower tip speed that reduces the overall noise produced by the system. OWTs have undergone significant computational analysis to uncover the underlying flow physics that can drive the system to high efficiencies for single wing oscillations. When two of these devices are placed in tandem configuration, i.e. one placed downstream of the other, they either can constructively or destructively interact. When constructive interactions occurred, they enhance the system efficiency to greater than that of two devices on their own. A new experimental design investigates the dependency of interaction modes on the pitch stiffness of the downstream wing. The experimental results demonstrated that interaction modes are functions of convective time scale and downstream wing pitch stiffness. Heterogeneous combinations of pitch stiffness exhibited constructive and destructive lock-in phenomena whereas the homogeneous combination exhibited only destructive interactions.
Guruswamy, G. P.; Goorjian, P. M.
1984-01-01
An efficient coordinate transformation technique is presented for constructing grids for unsteady, transonic aerodynamic computations for delta-type wings. The original shearing transformation yielded computations that were numerically unstable and this paper discusses the sources of those instabilities. The new shearing transformation yields computations that are stable, fast, and accurate. Comparisons of those two methods are shown for the flow over the F5 wing that demonstrate the new stability. Also, comparisons are made with experimental data that demonstrate the accuracy of the new method. The computations were made by using a time-accurate, finite-difference, alternating-direction-implicit (ADI) algorithm for the transonic small-disturbance potential equation.
Progress towards Steady State at Low Aspect Ratio on the National Spherical Torus Experiment (NSTX)
International Nuclear Information System (INIS)
Gates, D.A.; Menard, J.; Maingi, R.; Kaye, S.; Sabbagh, S.A.; Diem, S.; Wilson, J.R.; Bell, M.G.; Bell, R.E.; Ferron, J.; Fredrickson, E.D.; Kessel, C.E.; LeBlanc, B.P.; Levinton, F.; Manickam, J.; Mueller, D.; Raman, R.; Stevenson, T.; Stutman, D.; Taylor, G.; Tritz, K.; Yu, H.
2007-01-01
Modifications to the plasma control capabilities and poloidal field coils of the National Spherical Torus Experiment (NSTX) have enabled a significant enhancement in shaping capability which has led to the transient achievement of a record shape factor (S (triple b ond) q 95 (I p /aB t )) of ∼ 41 (MA m -1 T -1 ) simultaneous with a record plasma elongation of κ (triple b ond) b/a ∼ 3. This result was obtained using isoflux control and real-time equilibrium reconstruction. Achieving high shape factor together with tolerable divertor loading is an important result for future ST burning plasma experiments as exemplified by studies for future ST reactor concepts, as well as neutron producing devices, which rely on achieving high shape factors in order to achieve steady state operation while maintaining MHD stability. Statistical evidence is presented which demonstrates the expected correlation between increased shaping and improved plasma performance.
Minimizing scattering from antireflective surfaces replicated from low-aspect-ratio black silicon
DEFF Research Database (Denmark)
Christiansen, Alexander Bruun; Clausen, Jeppe; Mortensen, N. Asger
2012-01-01
The scattering properties of randomly structured antireflective black silicon polymer replica have been investigated. Using a two-step casting process, the structures can be replicated in Ormocomp on areas of up to 3 in. in diameter. Fourier analysis of scanning electron microscopy images...
Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations
International Nuclear Information System (INIS)
Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.
1986-06-01
Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost
UWMAK-II: a conceptual tokamak reactor design
International Nuclear Information System (INIS)
1975-10-01
This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high
Czech Academy of Sciences Publication Activity Database
Řípa, Milan; Křenek, Petr
2011-01-01
Roč. 17, č. 1 (2011), s. 32-34 ISSN 1210-4612 Institutional research plan: CEZ:AV0Z20430508 Keywords : fusion * tokamak * Compass * Golem * Institute of Plasma Physics AVCR v.v * NBI * diagnostics Subject RIV: BL - Plasma and Gas Discharge Physics
MAST: a Mega Amp Spherical Tokamak
International Nuclear Information System (INIS)
Darke, A.C.; Harbar, J.R.; Hay, J.H.; Hicks, J.B.; Hill, J.W.; McKenzie, J.S.; Morris, A.W.; Nightingale, M.P.S.; Todd, T.N.; Voss, G.M.; Watkins, J.R.
1995-01-01
The highly successful tight aspect ratio tokamak research pioneered on the START machine at Culham, together with the attractive possibilities of the concept, suggest a larger device should be considered. The design of a Mega Amp Spherical Tokamak is described, operating at much higher currents and over longer pulses than START and compatible with strong additional heating. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Komoshvili, K [Tel Aviv University, Ramat Aviv (Israel); Cuperman, S [Tel Aviv University, Ramat Aviv (Israel); Bruma, C [Tel Aviv University, Ramat Aviv (Israel)
2007-09-15
To assess the effect of antenna poloidal extension on fast waves-plasma interactions in pre-heated spherical tokamaks and, as a result, to assist the determination of optimal conditions for power deposition, we carried out a global, numerical investigation. Thus, we solved the steady-state full wave equations for Alfvenic modes in an inhomogeneous, non-uniformly magnetized, resistive, low aspect ratio tokamak plasma with appropriate consideration of boundary conditions; in this, processes such as wave propagation, reflection, transmission, absorption and mode conversion as well as mode-coupling(s) by plasma cross-section non-homogeneity generated waves were included. The results were analysed in terms of the directions of the current densities generated in the presence of up low field side or down high field side magnetic field gradient. Suitable antenna location and poloidal extension for maximum power deposition were determined.
International Nuclear Information System (INIS)
Komoshvili, K; Cuperman, S; Bruma, C
2007-01-01
To assess the effect of antenna poloidal extension on fast waves-plasma interactions in pre-heated spherical tokamaks and, as a result, to assist the determination of optimal conditions for power deposition, we carried out a global, numerical investigation. Thus, we solved the steady-state full wave equations for Alfvenic modes in an inhomogeneous, non-uniformly magnetized, resistive, low aspect ratio tokamak plasma with appropriate consideration of boundary conditions; in this, processes such as wave propagation, reflection, transmission, absorption and mode conversion as well as mode-coupling(s) by plasma cross-section non-homogeneity generated waves were included. The results were analysed in terms of the directions of the current densities generated in the presence of up low field side or down high field side magnetic field gradient. Suitable antenna location and poloidal extension for maximum power deposition were determined
A conceptual design of superconducting spherical tokamak reactor
International Nuclear Information System (INIS)
Nagayama, Yoshio; Shinya, Kichiro; Tanaka, Yasutoshi
2012-01-01
This paper presents a fusion reactor concept named 'JUST (Japanese Universities' Super Tokamak reactor)'. From the plasma confinement system to the power generation system is evaluated in this work. JUST design has features as follows: the superconducting magnet, the steady state operation with high bootstrap current fraction, the easy replacement of neutron damaged first wall, the high heat flux in the divertor, and the low cost (or high β). By winding the OH solenoid over the center stack of toroidal field coil, we have the low aspect ratio and the 80cm thick neutron shield to protect the superconducting center stack. JUST is designed by using the 0-D transport code under the assumption that the energy confinement time is 1.8 times of the IPB98(y,2) scaling. Main parameters are as follows: the major radius of 4.5m, the aspect ratio of 1.8, the elongation ratio of 2.5, the toroidal field of 2.36T, the plasma current of 18MA, the toroidal beta of 22%, the central electron and ion temperature of 15keV and the fusion thermal power of 2.4GW. By using the mercury heat exchanger and the steam turbine, the heat efficiency is 33% and the electric power is 0.74GW. (author)
On the design and role of passive stabilisation within the ST40 spherical tokamak
Buxton, P. F.; Asunta, O.; Gryaznevich, M. P.; Lockley, D.; McNamara, S.; Medvedev, S.; Ruiz de Villa Valdés, E.; Whitfield, G.; Wood, J. M.
2018-06-01
The position of passive stabilisation has been optimised for the low aspect ratio tokamak ST40. We find that passive stabilisation is most effective when conductors are placed near the plasma’s x-point, and the combined effect of having both inboard and outboard passive stabilisation significantly reduces the vertical instability growth rate. The growth rate can be further decreased by cooling the passive conductors down to 80 K. Two concepts for passive stabilisation are considered, passive plates and passive coils, and the relative advantages and disadvantages of each are discussed. Both concepts involve connecting the upper and lower conductors in an ‘anti-symmetric’ manner, which prevents large currents from being induced.
The role of the spherical tokamak in clarifying tokamak physics
International Nuclear Information System (INIS)
Morris, A.W.; Akers, R.J.; Connor, J.W.; Counsell, G.F.; Gryaznevich, M.P.; Hender, T.C.; Maddison, G.P.; Martin, T.J.; McClements, K.G.; Roach, C.M.; Robinson, D.C.; Sykes, A.; Valovic, M.; Wilson, H.R.; Fonck, R.J.; Gusev, V.; Kaye, S.M.; Majeski, R.; Peng, Y.-K.M.; Medvedev, S.; Sharapov, S.; Walsh, M.J.
1999-01-01
The spherical tokamak (ST) provides a unique environment in which to perform complementary and exacting tests of the tokamak physics required for a burning plasma experiment of any aspect ratio, while also having the potential for long-term fusion applications in its own right. New experiments are coming on-line in the UK (MAST), USA (NSTX, Pegasus), Russia (Globus-M), Brazil (ETE) and elsewhere, and the status of these devices will be reported, along with newly-analysed data from START. Those physics issues where the ST provides an opportunity to remove degeneracy in the databases or clarify one's understanding will be emphasized. (author)
International Nuclear Information System (INIS)
Cumyn, P.B.
A consortium of five organizations under the leadership of IREQ, the Institute de Recherche d'Hydro-Quebec has completed a conceptual design study for a tokamak device, and in January 1981 its construction was authorized with funding being provided principally by Hydro-Quebec and the National Research Council, as well as by the Ministre d'Education du Quebec and Natural Sciences and Engineering Research Council of Canada (NSERC). The device will form the focus of Canada's magnetic-fusion program and will be located in IREQ's laboratories in Varennes. Presently the machine layout is being finalized from the physics point of view and work has started on equipment design and specification. The Tokamak de Varennes will be an experimental device, the purpose of which is to study plasma and other fusion related phenomena. In particular it will study: 1. Plasma impurities and plasma/liner interaction; 2. Long pulse or quasi-continuous operation using plasma rampdown and eventually plasma current reversal in order to maintain the plasma; and 3. Advanced diagnostics
International Nuclear Information System (INIS)
Haines, M.G.
1984-01-01
The physical conditions required for breakeven in thermonuclear fusion are derived, and the early conceptual ideas of magnetic confinement and subsequent development are followed, leading to present-day large scale tokamak experiments. Confinement and diffusion are developed in terms of particle orbits, whilst magnetohydrodynamic stability is discussed from energy considerations. From these ideas are derived the scaling laws that determine the physical size and parameters of this fusion configuration. It becomes clear that additional heating is required. However there are currently several major gaps in our understanding of experiments; the causes of anomalous electron energy loss and the major current disruption, the absence of the 'bootstrap' current and what physics determines the maximum plasma pressure consistent with stability. The understanding of these phenomena is a major challenge to plasma physicists. (author)
Helical-tokamak hybridization concepts for compact configuration exploration and MHD stabilization
International Nuclear Information System (INIS)
Oishi, T.; Yamazaki, K.; Arimoto, H.; Baba, K.; Hasegawa, M.; Ozeki, H.; Shoji, T.; Mikhailov, M.I.
2010-11-01
To search for low-aspect-ratio torus systems, a lot of exotic confinement concepts are proposed so far historically. One of the authors previously proposed the tokamak-helical hybrid called TOKASTAR (Tokamak-Stellarator Hybrid) to improve the magnetic local shear near the bad curvature region. This is characterized by simple and compact coil systems with enough divertor space relevant to reactor designs. Based on this TOKASTAR concept, a toroidal mode number N=2 C (compact) -TOKASTAR machine (R - 35 mm) was constructed. The rotational transform of this compact helical configuration is rather small to confine hot ions, but can be utilized as a compact electron plasma machine for multi-purposes. The C-TOKASTAR has a pair of spherically winding helical coils and a pair of poloidal coils. Existence of magnetic surface and electron confinement property in C-TOKASTAR device were investigated by an electron-emission impedance method. Calculation of the particle orbit also supports that closed magnetic surface is formed in the cases that the ratio between poloidal and helical coil current is appropriate. Another aspect of the research using TOKASTAR configuration includes the evaluation of the effect of the outboard helical field application to tokamak plasmas. It is considered that outboard helical field has roles to assist the initiation of plasma current, to improve MHD stability, and so on. To check these roles, we made TOKASTAR-2 machine (R - 0.12 m, B - 1 kG) with ohmic heating central coil, eight toroidal field coils, a pair of vertical field coils and two outboard helical field coil segments. The electron cyclotron heating plasma start-up and plasma current disruption control experiments might be expected in this machine. Calculation of magnetic field line tracing has revealed that magnetic surface can be formed using additional outer helical coils. (author)
Particle control and plasma performance in the Lithium Tokamak eXperiment
Energy Technology Data Exchange (ETDEWEB)
Majeski, R.; Abrams, T.; Boyle, D.; Granstedt, E.; Hare, J.; Jacobson, C. M.; Kaita, R.; Kozub, T.; LeBlanc, B.; Lundberg, D. P.; Lucia, M.; Merino, E.; Schmitt, J.; Stotler, D. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Biewer, T. M.; Canik, J. M.; Gray, T. K.; Maingi, R.; McLean, A. G. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Kubota, S. [University of California at Los Angeles, Los Angeles, California 90095 (United States); and others
2013-05-15
The Lithium Tokamak eXperiment is a small, low aspect ratio tokamak [Majeski et al., Nucl. Fusion 49, 055014 (2009)], which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350 °C. Several gas fueling systems, including supersonic gas injection and molecular cluster injection, have been studied and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 ms. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 ms. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak—thin, evaporated, liquefied coatings of lithium—does not produce an adequately clean surface.
International Nuclear Information System (INIS)
Furth, H.P.
1984-10-01
The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT
International Nuclear Information System (INIS)
Rawls, J.M.
1979-10-01
An overall review of the tokamak program is given with particular emphasis upon developments over the past five years in the theoretical and experimental elements of the program. A summary of the key operating parameters for the principal tokamaks throughout the world is given. Also discussed are key issues in plasma confinement, plasma heating, and tokamak design
Tokamak and RFP ignition requirements
International Nuclear Information System (INIS)
Werley, K.A.
1991-01-01
A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab
International Nuclear Information System (INIS)
Reid, R.L.; Barrett, R.J.; Brown, T.G.
1985-03-01
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged
Tokamak devices: towards controlled fusion
International Nuclear Information System (INIS)
Trocheris, M.
1975-01-01
The Tokamak family is from Soviet Union. These devices were exclusively studied at the Kurchatov Institute in Moscow for more than ten years. The first occidental Tokamak started in 1970 at Princeton. The TFR (Tokamak Fontenay-aux-Roses) was built to be superior to the Russian T4. Tokamak future is now represented by the JET (Joint European Tokamak) [fr
MHD stability analysis of axisymmetric surface current model tokamaks close to the spheromak regime
International Nuclear Information System (INIS)
Honma, Toshihisa; Kaji, Ikuo; Fukai, Ichiro; Kito, Masafumi.
1984-01-01
In the toroidal coordinates, a stability analysis is presented for very low-aspect-ratio tokamaks with circular cross section which is described by a surface current model (SCM) of axisymmetric equilibria. The energy principle determining the stability of plasma is treated without any expansion of aspect ratio. Numerical results show that, owing to the occurrence of the non-axisymmetric (n=1) unstable modes, there exists no MHD-stable ideal SCM spheromak characterized by zero external toroidal vacuum field. Instead, a stable spheromak-type plasma which comes to the ideal SCM spheromak is provided by the configuration with a very weak external toroidal field. Close to the spheromak regime (1.0 1 aspect ratio< = 1.1), the minimum safety factor and the critical β-values increase mo notonically with aspect ratio decreasing from a large value, and curves of βsub(p) versus β in the marginal stability approach to an ideal SCM spheromak line βsub(p)=β. (author)
International Nuclear Information System (INIS)
Song, Yuntao; Wu, Weiyue; Du, Shijun
2014-01-01
Provides a systematic introduction to tokamaks in engineering mechanics. Includes design guides based on full mechanical analysis, which makes it possible to accurately predict load capacity and temperature increases. Presents comprehensive information on important design factors involving materials. Covers the latest advances in and up-to-date references on tokamak devices. Numerous examples reinforce the understanding of concepts and provide procedures for design. Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study of mechanical/fusion engineering with a general understanding of tokamak engineering mechanics.
Advanced commercial Tokamak optimization studies
International Nuclear Information System (INIS)
Whitley, R.H.; Berwald, D.H.; Gordon, J.D.
1985-01-01
Our recent studies have concentrated on developing optimal high beta (bean-shaped plasma) commercial tokamak configurations using TRW's Tokamak Reactor Systems Code (TRSC) with special emphasis on lower net electric power reactors that are more easily deployable. A wide range of issues were investigated in the search for the most economic configuration: fusion power, reactor size, wall load, magnet type, inboard blanket and shield thickness, plasma aspect ratio, and operational β value. The costs and configurations of both steady-state and pulsed reactors were also investigated. Optimal small and large reactor concepts were developed and compared by studying the cost of electricity from single units and from multiplexed units. Multiplexed units appear to have advantages because they share some plant equipment and have lower initial capital investment as compared to larger single units
International Nuclear Information System (INIS)
1991-05-01
discharges, production and self-organization of a turbulent plasma column in a spheromak (''SK-CG-1''), and (iv) a planned large-aspect ratio, high-beta tokamak (HBT-EP) experiment. Refs, figs and tabs
Song, Yuntao; Du, Shijun
2013-01-01
Tokamak Engineering Mechanics offers concise and thorough coverage of engineering mechanics theory and application for tokamaks, and the material is reinforced by numerous examples. Chapter topics include general principles, static mechanics, dynamic mechanics, thermal fluid mechanics and multiphysics structural mechanics of tokamak structure analysis. The theoretical principle of the design and the methods of the analysis for various components and load conditions are presented, while the latest engineering technologies are also introduced. The book will provide readers involved in the study
Advanced Tokamak Stability Theory
Zheng, Linjin
2015-03-01
The intention of this book is to introduce advanced tokamak stability theory. We start with the derivation of the Grad-Shafranov equation and the construction of various toroidal flux coordinates. An analytical tokamak equilibrium theory is presented to demonstrate the Shafranov shift and how the toroidal hoop force can be balanced by the application of a vertical magnetic field in tokamaks. In addition to advanced theories, this book also discusses the intuitive physics pictures for various experimentally observed phenomena.
Tokamak confinement scaling laws
International Nuclear Information System (INIS)
Connor, J.
1998-01-01
The scaling of energy confinement with engineering parameters, such as plasma current and major radius, is important for establishing the size of an ignited fusion device. Tokamaks exhibit a variety of modes of operation with different confinement properties. At present there is no adequate first principles theory to predict tokamak energy confinement and the empirical scaling method is the preferred approach to designing next step tokamaks. This paper reviews a number of robust theoretical concepts, such as dimensional analysis and stability boundaries, which provide a framework for characterising and understanding tokamak confinement and, therefore, generate more confidence in using empirical laws for extrapolation to future devices. (author)
International Nuclear Information System (INIS)
1986-04-01
This document contains the results of the IAEA Specialists' Meeting on Tokamak Concept Innovations held 13-17 January 1986 in Vienna. Although it is the most advanced fusion reactor concept the tokamak is not without its problems. Most of these problems should be solved within the ongoing R and D studies for the next generation of tokamaks. Emphasis for this meeting was placed on innovations that would lead to substantial improvements in a tokamak reactor, even if they involved a radical departure from present thinking
International Nuclear Information System (INIS)
Edelbaum, T.N.; Serben, S.; Var, R.E.
1976-01-01
A computer model of a tokamak experimental power reactor and its control system is being constructed. This simulator will allow the exploration of various open loop and closed loop strategies for reactor control. This paper provides a brief description of the simulator and some of the potential control problems associated with this class of tokamaks
Peterson, Ethan; Anderson, Jay; Clark, Mike; Egedal, Jan; Endrizzi, Douglass; Flanagan, Ken; Harvey, Robert; Lynn, Jacob; Milhone, Jason; Wallace, John; Waleffe, Roger; Mirnov, Vladimir; Forest, Cary
2017-10-01
Equilibrium reconstructions of rotating magnetospheres in the lab are computed using a user-friendly extended Grad-Shafranov solver written in Python and various magnetic and kinetic measurements. The stability of these equilibria are investigated using the NIMROD code with two goals: understand the onset of the classic ``wobble'' in the heliospheric current sheet and demonstrating proof-of-principle for a laboratory source of high- β turbulence. Using the same extended Grad-Shafranov solver, equilibria for an axisymmetric, non-paraxial magnetic mirror are used as a design foundation for a high-field magnetic mirror neutron source. These equilibria are numerically shown to be stable to the m=1 flute instability, with higher modes likely stabilized by FLR effects; this provides stability to gross MHD modes in an axisymmetric configuration. Numerical results of RF heating and neutral beam injection (NBI) from the GENRAY/CQL3D code suite show neutron fluxes promising for medical radioisotope production as well as materials testing. Synergistic effects between NBI and high-harmonic fast wave heating show large increases in neutron yield for a modest increase in RF power. work funded by DOE, NSF, NASA.
Performance and stability limits at near-unity aspect ratio in the Pegasus Toroidal Experiment
International Nuclear Information System (INIS)
Fonck, R.J.
2002-01-01
The Pegasus Toroidal Experiment is a mid-sized extremely-low aspect ratio (A) spherical torus (ST). It has the dual roles of exploring limits of ST behavior as A approaches 1 and studying the physics of ST plasmas in the tokamak-spheromak overlap regime. Major parameters are R 0.25 - 0.45 m, A 1.1 - 1.4, I p ≤ 0.15MA, and B t p =aB t is similar to that observed for NBI-heated START discharges. Achievable plasma current apparently is subject to a 'soft' limit of I p =I t f ≤ 1. Access to higher-current plasmas appears to be restricted by the appearance of large internal MHD activity, including m/n=2/1 and 3/2 modes. Recent experiments have begun to access ideal stability limits, with disruptions observed as q 95 approaches 5, in agreement with numerical predictions. (author)
System assessment of helical reactors in comparison with tokamaks
International Nuclear Information System (INIS)
Yamazaki, K.; Imagawa, S.; Muroga, T.; Sagara, A.; Okamura, S.
2002-10-01
A comparative assessment of tokamak and helical reactors has been performed using equivalent physics/engineering model and common costing model. Higher-temperature plasma operation is required in tokamak reactors to increase bootstrap current fraction and to reduce current-drive (CD) power. In helical systems, lower-temperature operation is feasible and desirable to reduce helical ripple transport. The capital cost of helical reactor is rather high, however, the cost of electricity (COE) is almost same as that of tokamak reactor because of smaller re-circulation power (no CD power) and less-frequent blanket replacement (lower neutron wall loading). The standard LHD-type helical reactor with 5% beta value is economically equivalent to the standard tokamak with 3% beta. The COE of lower-aspect ratio helical reactor is on the same level of high-β N tokamak reactors. (author)
International Nuclear Information System (INIS)
Wesson, John; Campbell, D.J.; Connor, J.W.
1997-01-01
It is interesting to recall the state of tokamak research when the first edition of this book was written. My judgement of the level of real understanding at that time is indicated by the virtual absence of comparisons of experiment with theory in that edition. The need then was for a 'handbook' which collected in a single volume the concepts and models which form the basis of everyday tokamak research. The experimental and theoretical endeavours of the subsequent decade have left almost all of this intact, but have brought a massive development of the subject. Firstly, there are now several areas where the experimental behaviour is described in terms of accepted theory. This is particularly true of currents parallel to the magnetic field, and of the stability limitations on the plasma pressure. Next there has been the research on large tokamaks, hardly started at the writing of the first edition. Now our thinking is largely based on the results from these tokamaks and this work has led to the long awaited achievement of significant amounts of fusion power. Finally, the success of tokamak research has brought us face to face with the problems involved in designing and building a tokamak reactor. The present edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes an account of the advances outlined above. (Author)
International Nuclear Information System (INIS)
Baker, C.C.
1981-01-01
This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features
International Nuclear Information System (INIS)
Murray, J.G.; Gorker, G.E.
1985-01-01
Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage
International Nuclear Information System (INIS)
Bickerton, R.J.
1977-01-01
The survey covers the following topics:- Introduction and history of tokamak research; review of tokamak apparatus, existing and planned; remarks on measurement techniques and their limitations; main results in terms of electron and ion temperatures, plasma density, containment times, etc. Empirical scaling; range of operating densities; impurities, origin, behaviour and control (including divertors); data on fluctuations and instabilities in tokamak plasmas; data on disruptive instabilities; experiments on shaped cross-sections; present experimental evidence on β limits; auxiliary heating; experimental and theoretical problems for the future. (author)
Energy Technology Data Exchange (ETDEWEB)
Murray, J.G.; Gorker, G.E.
1985-01-01
Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.
Start-up of spherical tokamak without a center solenoid
International Nuclear Information System (INIS)
Maekawa, Takashi; Nagata, Masayoshi
2012-01-01
For low-aspect tokamak reactors, spherical tokamak reactors, ST-type FESF/CTFs, it is essential to remove or minimize a central solenoid (CS). Even with the minimized CS, non-inductive start up of the plasma current is required. Rapid increase in the spontaneous plasma current at the final stage of current start-up drives ignition. At the initial stage, formation of plasma and magnetic surfaces are required. As non-inductive plasma start-up scenarios, ECH/ECCD, LHCD, HHFW, DC HELICITY injection, plasma merging and NBI have been studied. In the present article, the present status and future prospect of experimental and theoretical works on these subjects. (author)
Tokamak engineering test reactor
International Nuclear Information System (INIS)
Conn, R.W.; Jassby, D.L.
1975-07-01
The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
1998-12-31
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
1997-12-31
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author) 55 refs., 40 figs.
Tokamak simulation code manual
International Nuclear Information System (INIS)
Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won
1995-01-01
The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs
HESTER: a hot-electron superconducting tokamak experimental reactor at M.I.T
International Nuclear Information System (INIS)
Schultz, J.H.; Montgomery, D.B.
1983-04-01
HESTER is an experimental tokamak, designed to resolve many of the central questions in the tokamak development program in the 1980's. It combines several unique features with new perspectives on the other major tokamak experiments scheduled for the next decade. The overall objectives of HESTER, in rough order of their presently perceived importance, are the achievement of reactor-like wall-loadings and plasma parameters for long pulse periods, determination of a good, reactor-relevant method of steady-state or very long pulse tokamak current drive, duplication of the planned very high temperature neutral injection experiments using only radio frequency heating, a demonstration of true steady-state tokamak operation, integration of a high-performance superconducting magnet system into a tokamak experiment, determination of the best methods of long term impurity control, and studies of transport and pressure limits in high field, high aspect ratio tokamak plasmas. These objectives are described
The residual zonal dynamics in a toroidally rotating tokamak
International Nuclear Information System (INIS)
Zhou Deng
2015-01-01
Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In this presentation, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved to give the expression of residual zonal flows with arbitrary rotating velocity. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the previous simulation result for high aspect ratio tokamaks. (author)
Joint research using small tokamaks
International Nuclear Information System (INIS)
Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Oost, G. van
2005-01-01
Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive co-ordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Co-ordinated Research Project is presented. (author)
Joint research using small tokamaks
International Nuclear Information System (INIS)
Gryaznevich, M.P.; Bosco, E. Del; Malaquias, A.; Mank, G.; Oost, G. van; He, Yexi; Hegazy, H.; Hirose, A.; Hron, M.; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.
2005-01-01
Small tokamaks have an important role in fusion research. More than 40 small tokamaks are operational. Research on small tokamaks has created a scientific basis for the scaling-up to larger tokamaks. Well-known scientific and engineering schools, which are now determining the main directions of fusion science and technology, have been established through research on small tokamaks. Combined efforts within a network of small and medium size tokamaks will further enhance the contribution of small tokamaks. A new concept of interactive coordinated research using small tokamaks in the mainstream fusion science areas, in testing of new diagnostics, materials and technologies as well as in education, training and broadening of the geography of fusion research in the scope of the IAEA Coordinated Research Project, is presented
Integrated predictive modeling simulations of the Mega-Amp Spherical Tokamak
International Nuclear Information System (INIS)
Nguyen, Canh N.; Bateman, Glenn; Kritz, Arnold H.; Akers, Robert; Byrom, Calum; Sykes, Alan
2002-01-01
Integrated predictive modeling simulations are carried out using the BALDUR transport code [Singer et al., Comput. Phys. Commun. 49, 275 (1982)] for high confinement mode (H-mode) and low confinement mode (L-mode) discharges in the Mega-Amp Spherical Tokamak (MAST) [Sykes et al., Phys. Plasmas 8, 2101 (2001)]. Simulation results, obtained using either the Multi-Mode transport model (MMM95) or, alternatively, the mixed-Bohm/gyro-Bohm transport model, are compared with experimental data. In addition to the anomalous transport, neoclassical transport is included in the simulations and the ion thermal diffusivity in the inner third of the plasma is found to be predominantly neoclassical. The sawtooth oscillations in the simulations radially spread the neutral beam injection heating profiles across a broad sawtooth mixing region. The broad sawtooth oscillations also flatten the central temperature and electron density profiles. Simulation results for the electron temperature and density profiles are compared with experimental data to test the applicability of these models and the BALDUR integrated modeling code in the limit of low aspect ratio toroidal plasmas
Advanced commercial tokamak study
International Nuclear Information System (INIS)
Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.
1985-12-01
Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs
International Nuclear Information System (INIS)
Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.
2009-01-01
Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.
Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation
International Nuclear Information System (INIS)
Riedel, K.S.
1989-03-01
This is a compendium of three separate articles on the statistical analysis of tokamak transport. The first article is an expository introduction to advanced statistics and scaling laws. The second analyzes two important problems of tokamak data---colinearity and tokamak to tokamak variation in detail. The third article generalizes the Swamy random coefficient model to the case of degenerate matrices. Three papers have been processed separately
International Nuclear Information System (INIS)
Anon.
1988-01-01
The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. The experiment, soon to be operational, provides an opportunity to study dense plasmas heated by powers unprecedented in the electron-cyclotron frequency range required by the especially high magnetic fields used with the MTX and needed for reactors. 1 references, 5 figures, 3 tables
International Nuclear Information System (INIS)
Wootton, A.J.
1993-04-01
This progress report covers the period from November 1, 1990 to April 30, 1993. During that period, TEXT was operated as a circular tokamak with a material limiter. It was devoted to the study of basic plasma physics, in particular to study of fluctuations, turbulence, and transport. The purpose is to operate and maintain TEXT Upgrade as a complete facility for applied tokamak physics, specifically to conduct a research program under the following main headings: (1) to elucidate the mechanisms of working gas, impurity, and thermal transport in tokamaks, in particular to understand the role of turbulence; (2) to study physics of the edge plasma, in particular the turbulence; (3) to study the physics or resonant magnetic fields (ergodic magnetic divertors, intra island pumping); and (4) to study the physics of electron cyclotron heating (ECRH). Results of studies in each of these areas are reported
Magnetic ''islandography'' in tokamaks
International Nuclear Information System (INIS)
Callen, J.D.; Waddell, B.V.; Hicks, H.R.
1978-09-01
Tearing modes are shown to be responsible for most of the experimentally observed macroscopic behavior of tokamak discharges. The effects of these collective magnetic perturbations on magnetic topology and plasma transport in tokamaks are shown to provide plausible explanations for: internal disruptions (m/n = 1); Mirnov oscillations (m/n = 2,3...); and major disruptions (coupling of 2/1-3/2 modes). The nonlinear evolution of the tearing modes is followed with fully three-dimensional computer codes. The effects on plasma confinement of the magnetic islands or stochastic field lines induced by the macroscopic tearing modes are discussed and compared with experiment. Finally, microscopic magnetic perturbations are shown to provide a natural model for the microscopic anomalous transport processes in tokamaks
Accelerator technology in tokamaks
International Nuclear Information System (INIS)
Kustom, R.L.
1977-01-01
This article presents the similarities in the technology required for high energy accelerators and tokamak fusion devices. The tokamak devices and R and D programs described in the text represent only a fraction of the total fusion program. The technological barriers to producing successful, economical tokamak fusion power plants are as many as the plasma physics problems to be overcome. With the present emphasis on energy problems in this country and elsewhere, it is very likely that fusion technology related R and D programs will vigorously continue; and since high energy accelerator technology has so much in common with fusion technology, more scientists from the accelerator community are likely to be attracted to fusion problems
International Nuclear Information System (INIS)
Steiner, D.; Embrechts, M.
1990-07-01
This is a status report on technical progress relative to the tasks identified for the fifth year of Grant No. FG02-85-ER52118. The ARIES tokamak reactor study is a multi-institutional effort to develop several visions of the tokamak as an attractive fusion reactor with enhanced economic, safety, and environmental features. The ARIES study is being coordinated by UCLA and involves a number of institutions, including RPI. The RPI group has been pursuing the following areas of research in the context of the ARIES-I design effort: MHD equilibrium and stability analyses; plasma-edge modeling and blanket materials issues. Progress in these areas is summarized herein
Internal disruption in tokamaks
International Nuclear Information System (INIS)
Kuvshinov, B.N.; Savrukhin, P.V.
1990-01-01
A review of results of experimental and theoretical investigations of internal disruption in tokamaks is given. Specific features of various types of saw-tooth oscillations are described and their classification is performed. Theoretical models of the process of development of internal disruption instability are discussed. Effect of internal disruption on parameters of plasma, confined in tokamak, is considered. Scalings of period and amplitude of saw-tooth oscillations, as well as version radius are presented. Different methods for stabilizing instability of internal disruption are described
International Nuclear Information System (INIS)
Unterberg, Bernhard; Samm, Ulrich
2004-01-01
An overview is given of recent results obtained in tokamak devices. We introduce basic confinement scenarios as L-mode, H-mode and plasmas with an internal transport barrier and discuss methods for profile control. Important findings in DT-experiments at JET as α-particle heating are described. Methods for power exhaust like plasma regimes with a radiating mantle and radiative divertor scenarios are discussed. The overall impact of plasma edge conditions on the general plasma performance in tokamaks is illustrated by describing the impact of wall conditions on confinement and the edge operational diagram of H-mode plasmas
Internal disruptions in tokamaks
International Nuclear Information System (INIS)
Kuvshinov, B.N.; Savrukhin, P.V.
1990-01-01
Experimental and theoretical studies of the phenomenon of internal disruptions in tokamaks are reviewed. A classification scheme is introduced and the features of different types of sawtooth oscillations are described. A theoretical model for the development of the internal disruption instability is discussed. The effect of internal disruptions on the parameters of plasma confined in tokamaks is discussed. Scaling laws for the period and amplitude of sawtooth oscillations, as well as for the inversion radius, are presented. Different methods of stabilizing the internal disruption instability are described
Accessibility of second regions of stability in tokamaks
Energy Technology Data Exchange (ETDEWEB)
Manickam, J.
1985-12-01
Second regions of stability to the ideal ballooning modes have been shown to exist in large-aspect-ratio circular and small-aspect-ratio bean-shaped tokamaks. We report on the existence of these second stability regions in finite-aspect-ratio dee-shaped tokamaks. We also report on the discovery of a second-stable region with respect to the n = 1 external kink mode in a bean-shaped plasma. The role of the shear and current profile in determining these regions of parameter space are discussed. 13 refs., 6 figs.
Accessibility of second regions of stability in tokamaks
International Nuclear Information System (INIS)
Manickam, J.
1985-12-01
Second regions of stability to the ideal ballooning modes have been shown to exist in large-aspect-ratio circular and small-aspect-ratio bean-shaped tokamaks. We report on the existence of these second stability regions in finite-aspect-ratio dee-shaped tokamaks. We also report on the discovery of a second-stable region with respect to the n = 1 external kink mode in a bean-shaped plasma. The role of the shear and current profile in determining these regions of parameter space are discussed. 13 refs., 6 figs
The basics of spherical tokamaks and progress in European research
International Nuclear Information System (INIS)
Gusev, V K; Alladio, F; Morris, A W
2003-01-01
When the aspect ratio of a tokamak (A = R/a) decreases significantly, there is a transformation of the well studied tokamak toroidal magnetic configuration into the spherical tokamak (ST) configuration. This configuration has high natural plasma elongation and triangularity and other unique equilibrium and stability properties of ST configuration, which are discussed in this paper. European research into ST physics is well advanced in spite of the young age of this branch of fusion science. An overview of selected experimental and theoretical results obtained at Ioffe, Culham and Frascati is given with the emphasis on their complementarity and links to the main stream of tokamak research, such as ITER. An outline of the basic ST advantages and the potential of ST research for new insights into magnetic confinement is also given. More detailed descriptions of recent advances in ST theory and experiment may be found in the invited papers by Akers and Ono in the proceedings of this conference
International Nuclear Information System (INIS)
Dory, R.A.; Berger, D.P.; Charlton, L.A.; Hogan, J.T.; Munro, J.K.; Nelson, D.B.; Peng, Y.K.M.; Sigmar, D.J.; Strickler, D.J.
1978-01-01
MHD equilibrium, stability, and transport calculations are made to study the accessibility and behavior of ''high beta'' tokamak plasmas in the range β approximately 5 to 15 percent. For next generation devices, beta values of at least 8 percent appear to be accessible and stable if there is a conducting surface nearby
Sawtooth phenomena in tokamaks
International Nuclear Information System (INIS)
Kuvshinov, B.N.; Savrukhin, P.V.
1989-01-01
A review of experimental and theoretical investigaions of sawtooth phenomena in tokamaks is presented. Different types of sawtooth oscillations, scaling laws and methods of interanl disruption stabilization are described. Theoretical models of the sawtooth instability are discussed. 122 refs.; 4 tabs
International Nuclear Information System (INIS)
Pare, V.K.
1983-01-01
Calculations with several different computer codes based on the resistive MHD equations have shown that (m = 1, n = 1) tearing modes in tokamak plasmas grow by magnetic reconnection. The observable behavior predicted by the codes has been confirmed in detail from the waveforms of signals from x-ray detectors and recently by x-ray tomographic imaging
International Nuclear Information System (INIS)
1993-01-01
This document consists of a collection of papers presented at the IAEA Technical Committee Meeting on Research Using Small Tokamaks. It contains 22 papers on a wide variety of research aspects, including diagnostics, design, transport, equilibrium, stability, and confinement. Some of these papers are devoted to other concepts (stellarators, compact tori). Refs, figs and tabs
International Nuclear Information System (INIS)
1991-01-01
The technical reports contained in this collection of papers on research using small tokamaks fall into four main categories, i.e., (i) experimental work (heating, stability, plasma radial profiles, fluctuations and transport, confinement, ultra-low-q tokamaks, wall physics, a.o.), (ii) diagnostics (beam probes, laser scattering, X-ray tomography, laser interferometry, electron-cyclotron absorption and emission systems), (iii) theory (strong turbulence, effects of heating on stability, plasma beta limits, wave absorption, macrostability, low-q tokamak configurations and bootstrap currents, turbulent heating, stability of vortex flows, nonlinear islands growth, plasma-drift-induced anomalous transport, ergodic divertor design, a.o.), and (iv) new technical facilities (varistors applied to establish constant current and loop voltage in HT-6M), lower-hybrid-current-drive systems for HT-6B and HT-6M, radio-frequency systems for HT-6M ICR heating experimentation, and applications of fiber optics for visible and vacuum ultraviolet radiation detection as applied to tokamaks and reversed-field pinches. A total number of 51 papers are included in the collection. Refs, figs and tabs
International Nuclear Information System (INIS)
Wootton, A.J.
1990-04-01
This paper discusses the following work on the text tokamak: data systems; particle confinement; impurity transport; plasma rotation; runaway electrons; electron cyclotron heating; FIR system; transient transport; internal turbulence; edge turbulence; ion temperature; EML experiments; impurity pellet experiments; MHD experiments and analysis; TEXT Upgrade; and Upgrade diagnostics
The ARIES tokamak fusion reactor study
International Nuclear Information System (INIS)
Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.
1989-01-01
The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab
Energy Technology Data Exchange (ETDEWEB)
Stott, Peter [JET, UK (United Kingdom)
1998-10-01
The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)
International Nuclear Information System (INIS)
Stott, Peter
1998-01-01
The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)
Control and Data Acquisition for the Spherical Tokamak MEDUSA-CR
Soto, Christian; Gonzalez, Jeferson; Carvajal, Johan; Ribeiro, Celso
2013-10-01
The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R loan to our laboratory via NI-Costa Rica. The interface with the energy, gas fueling, and security systems are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.
Present status of Tokamak research
International Nuclear Information System (INIS)
Basu, Jayanta
1991-01-01
The scenario of thermonuclear fusion research is presented, and the tokamak which is the most promising candidate as a fusion reactor is introduced. A brief survey is given of the most noteworthy tokamaks in the global context, and fusion programmes relating to Next Step devices are outlined. Supplementary heating of tokamak plasma by different methods is briefly reviewed; the latest achievements in heating to fusion temperatures are also reported. The progress towards the high value of the fusion product necessary for ignition is described. The improvement in plasma confinement brought about especially by the H-mode, is discussed. The latest situation in pushing up Β for increasing the efficiency of a tokamak is elucidated. Mention is made of the different types of wall treatment of the tokamak vessel for impurity control, which has led to a significant improvement in tokamak performance. Different methods of current drive for steady state tokamak operation are reviewed, and the issue of current drive efficiency is addressed. A short resume is given of the various diagnostic methods which are employed on a routine basis in the major tokamak centres. A few diagnostics recently developed or proposed in the context of the advanced tokamaks as well as the Next Step devices are indicated. The important role of the interplay between theory, experiment and simulation is noted, and the areas of investigation requiring concerted effort for further progress in tokamak research are identified. (author). 17 refs
Stability of Tokamaks with respect to slip motions
International Nuclear Information System (INIS)
Rebhan, E.; Salat, A.
1976-06-01
Using the energy principle in Tokamaks we investigate a class of perturbations which, if unstable, cannot be stabilized by the toroidal main field. On the assumptions of usual Tokamak ordering and in the limit of infinite aspect ratio, these perturbations are shown to be minimizing among all axisymmetric perturbations. In the case of finite aspect ratio, a detailed stability analysis is carried out using a constant pressure surface current model with elliptic, triangular or rectangular plasma cross-section. Definite stabilization by toroidal effects and by beta poloidal is demonstrated. (orig.) [de
The calculation of Tritium burnup in Tokamaks
International Nuclear Information System (INIS)
Bittoni, E.; Haegi, M.
1987-01-01
In a deuterium plasma tokamak, the contained fusion-produced tritons are supposed to be decelerated down to thermalization according to classical Coulomb scattering. A fraction of these fast tritons undergoes the DT fusion reaction producing 14.1 MeV neutrons. It is thus possible to get information on the confinement of these fast tritons by comparing the measured and the calculated ratio of the 14.1 MeV to the 2.45 MeV neutron flux. This report describes the calculation of this flux ratio by means of a numerical Monte Carlo-like code
International Nuclear Information System (INIS)
Weldon, D.M.; Murray, J.G.
1983-01-01
Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor
Tokamak fusion reactor exhaust
International Nuclear Information System (INIS)
Harrison, M.F.A.; Harbour, P.J.; Hotston, E.S.
1981-08-01
This report presents a compilation of papers dealing with reactor exhaust which were produced as part of the TIGER Tokamak Installation for Generating Electricity study at Culham. The papers are entitled: (1) Exhaust impurity control and refuelling. (2) Consideration of the physical problems of a self-consistent exhaust and divertor system for a long burn Tokamak. (3) Possible bundle divertors for INTOR and TIGER. (4) Consideration of various magnetic divertor configurations for INTOR and TIGER. (5) A appraisal of divertor experiments. (6) Hybrid divertors on INTOR. (7) Refuelling and the scrape-off layer of INTOR. (8) Simple modelling of the scrape-off layer. (9) Power flow in the scrape-off layer. (10) A model of particle transport within the scrape-off plasma and divertor. (11) Controlled recirculation of exhaust gas from the divertor into the scrape-off plasma. (U.K.)
Energy Technology Data Exchange (ETDEWEB)
White, R B [Princeton Univ., NJ (USA). Plasma Physics Lab.
1989-01-01
The book covers the consequences of ideal and resistive magnetohydrodynamics, these theories being responsible for most of what is well understood regarding the physics of tokamak discharges. The focus is on the description of equilibria, the linear and nonlinear theory of large scale modes, and single particle guiding center motion, including simple neoclassical effects. modern methods of general magnetic coordinates are used, and the student is introduced to the onset of chaos in Hamiltonian systems in the discussion of destruction of magnetic surfaces. Much of the book is devoted to the description of the limitations placed on tokamak operating parameters given by ideal and resistive modes, and current ideas about how to extend and optimize these parameters. (author). refs.; figs.
Axisymmetric tokamak scapeoff transport
International Nuclear Information System (INIS)
Singer, C.E.; Langer, W.D.
1982-08-01
We present the first self-consistent estimate of the magnitude of each term in a fluid treatment of plasma transport for a plasma lying in regions of open field lines in an axisymmetric tokamak. The fluid consists of a pure hydrogen plasma with sources which arise from its interaction with neutral hydrogen atoms. The analysis and results are limited to the high collisionality regime, which is optimal for a gaseous neutralizer divertor, or to a cold plasma mantle in a tokamak reactor. In this regime, both classical and neoclassical transport processes are important, and loss of particles and energy by diamagnetic flow are also significant. The prospect of extending the analysis to the lower collisionality regimes encountered in many existing experiments is discussed
International Nuclear Information System (INIS)
Tendler, M.
1984-06-01
The energy loss from a tokamak plasma due to neutral hydrogen radiation and recycling is of great importance for the energy balance at the periphery. It is shown that the requirement for thermal equilibrium implies a constraint on the maximum attainable edge density. The relation to other density limits is discussed. The average plasma density is shown to be a strong function of the refuelling deposition profile. (author)
International Nuclear Information System (INIS)
Conn, R.W.
1984-05-01
Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6MW of auxiliary neutral beam heating. Experiments have also been done with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a region may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this Z-mode of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described
Modular tokamak magnetic system
International Nuclear Information System (INIS)
Yang, T.F.
1988-01-01
This patent describes a tokamak reactor including a vacuum vessel, toroidal confining magnetic field coils disposed concentrically around the minor radius of the vacuum vessel, and poloidal confining magnetic field coils, an ohmic heating coil system comprising at least one magnetic coil disposed concentrically around a toroidal field coil, wherein the magnetic coil is wound around the toroidal field coil such that the ohmic heating coil enclosed the toroidal field coil
International Nuclear Information System (INIS)
Conn, R.W.; California Univ., Los Angeles
1984-01-01
Recent experiments with a scoop limiter without active internal pumping have been carried out in the PDX tokamak with up to 6 MW of auxiliary neutral beam heating. Experiments have also been performed with a rotating head pump limiter in the PLT tokamak in conjunction with RF plasma heating. Extensive experiments have been done in the ISX-B tokamak and first experiments have been completed with the ALT-I limiter in TEXTOR. The pump limiter modules in these latter two machines have internal getter pumping. Experiments in ISX-B are with ohmic and auxiliary neutral beam heating. The results in ISX-B and TEXTOR show that active density control and particle removal is achieved with pump limiters. In ISX-B, the boundary layer (or scrape-off layer) plasma partially screens the core plasma from gas injection. In both ISX-B and TEXTOR, the pressure internal to the module scales linearly with plasma density but in ISX-B, with neutral beam injection, a nonlinear increase is observed at the highest densities studied. Plasma plugging is the suspected cause. Results from PDX suggest that a regime may exist in which core plasma energy confinement improves using a pump limiter during neutral beam injection. Asymmetric radial profiles and an increased edge electron temperature are observed in discharges with improved confinement. The injection of small amounts of neon into ISX-B has more clearly shown an improved electron core energy confinement during neutral beam injection. While carried out with a regular limiter, this 'Z-mode' of operation is ideal for use with pump limiters and should be a way to achieve energy confinement times similar to values for H-mode tokamak plasmas. The implication of all these results for the design of a reactor pump limiter is described. (orig.)
TPX tokamak construction management
International Nuclear Information System (INIS)
Knutson, D.; Kungl, D.; Seidel, P.; Halfast, C.
1995-01-01
A construction management contract normally involves the acquisition of a construction management firm to assist in the design, planning, budget conformance, and coordination of the construction effort. In addition the construction management firm acts as an agent in the awarding of lower tier contracts. The TPX Tokamak Construction Management (TCM) approach differs in that the construction management firm is also directly responsible for the assembly and installation of the tokamak including the design and fabrication of all tooling required for assembly. The Systems Integration Support (SIS) contractor is responsible for the architect-engineering design of ancillary systems, such as heating and cooling, buildings, modifications and site improvements, and a variety of electrical requirements, including switchyards and >4kV power distribution. The TCM will be responsible for the procurement of materials and the installation of the ancillary systems, which can either be performed directly by the TCM or subcontracted to a lower tier subcontractor. Assurance that the TPX tokamak is properly assembled and ready for operation when turned over to the operations team is the primary focus of the construction management effort. To accomplish this a disciplined constructability program will be instituted. The constructability effort will involve the effective and timely integration of construction expertise into the planning, component design, and field operations. Although individual component design groups will provide liaison during the machine assembly operations, the construction management team is responsible for assembly
Non-inductively driven tokamak plasmas at near-unity βt in the Pegasus toroidal experiment
Reusch, J. A.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Perry, J. M.; Pierren, C.; Rhodes, A. T.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Weberski, J. D.
2018-05-01
A major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓi, high elongation κ, and high toroidal and normalized beta ( βt and βN) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓi. The low aspect ratio ( R0/a ˜1.2 ) of Pegasus allows access to high κ and high normalized plasma currents ( IN=Ip/a BT>14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high βt plasmas. Equilibrium analyses indicate that βt up to ˜100% is achieved. These high βt discharges disrupt at the ideal no-wall β limit at βN˜7.
Sheffield, J.
1981-08-01
For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.
Magnetic confinement experiment -- 1: Tokamaks
International Nuclear Information System (INIS)
Goldston, R.J.
1994-01-01
This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization
International Nuclear Information System (INIS)
Alladio, F.
1985-01-01
A short summary of the present status of experimental observations, theoretical ideas and understanding of the density limit in tokamaks is presented. It is the result of the discussion that was held on this topic at the 4th European Tokamak Workshop in Copenhagen (December 4th to 6th, 1985). 610 refs
International Nuclear Information System (INIS)
Rogister, A L
2004-01-01
John Wesson's well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author's task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2-9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to
MHD stability of an almost circular tokamak
International Nuclear Information System (INIS)
Roy, A.
1990-10-01
In a tokamak, the ratio β between the plasma pressure and that of the magnetic field is limited by the appearance of instabilities. The magnetic field in a tokamak reactor will always be limited by technological constraints. It is therefore crucial to know what factors have an effect on the β limit, since a zero resistivity plasma fluid model allows for theoretical reproduction of the β limits observed experimentally. Theoretical studies have shown that the distributions of pressure and current density may have a substantial effect on the β limit. The effect of the current density and pressure distributions on the β limit has been studied for tokamak with a circular core section. The best results are obtained when the current density is concentrated in the centre of the section and is nil at the periphery. But the second region of stability against ballooning modes cannot be obtained in a circular tokamak owing to the destabilisation of the universal modes. This study was then extended to the stability of plasmas the section of which is almost circular and has a point of reflection. Such configurations are vital for fusion since they allow systems in which the confinement time does not deteriorate with an increase in the additional heating power. The β limit was calculated for different positions of the reflection point. The results show that when it is displaced from the interior towards the exterior of the torus, the stability of the overall modes is progressively improved until it is vertical. But if the point of reflection is further displaced from this vertical position towards the exterior of the torus, localised modes close to the edge of the plasma are destabilised and bring about a drop in the β limit. (author) figs., tabs., 80 refs
The tokamak as a neutron source
International Nuclear Information System (INIS)
Hendel, H.W.; Jassby, D.L.
1989-11-01
This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs
Tokamak instrumentation and controls
International Nuclear Information System (INIS)
Becraft, W.R.; Bettis, E.S.; Houlberg, W.A.; Onega, R.J.; Stone, R.S.
1979-02-01
The three areas of study emphasis to date are: (1) Physics implications for controls, (2) Computer simulation, and (3) Shutdown/aborts. This document reports on the FY 78 efforts (the first year of these studies) to address these problems. Transient scenario options for the startup of a tokamak are developed, and the implications for the control system are discussed. This document also presents a hybrid computer simulation (analog and digital) of the Impurity Study Experiment (ISX-B) which is now being used for corroborative controls investigations. The simulation will be expanded to represent a TNS/ETF machine
Demonstration tokamak power plant
International Nuclear Information System (INIS)
Abdou, M.; Baker, C.; Brooks, J.; Ehst, D.; Mattas, R.; Smith, D.L.; DeFreece, D.; Morgan, G.D.; Trachsel, C.
1983-01-01
A conceptual design for a tokamak demonstration power plant (DEMO) was developed. A large part of the study focused on examining the key issues and identifying the R and D needs for: (1) current drive for steady-state operation, (2) impurity control and exhaust, (3) tritium breeding blanket, and (4) reactor configuration and maintenance. Impurity control and exhaust will not be covered in this paper but is discussed in another paper in these proceedings, entitled Key Issues of FED/INTOR Impurity Control System
Maximum entropy tokamak configurations
International Nuclear Information System (INIS)
Minardi, E.
1989-01-01
The new entropy concept for the collective magnetic equilibria is applied to the description of the states of a tokamak subject to ohmic and auxiliary heating. The condition for the existence of steady state plasma states with vanishing entropy production implies, on one hand, the resilience of specific current density profiles and, on the other, severe restrictions on the scaling of the confinement time with power and current. These restrictions are consistent with Goldston scaling and with the existence of a heat pinch. (author)
International Nuclear Information System (INIS)
Barbosa, Luis Filipe F.P.W.; Bosco, Edson del.
1994-01-01
This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs
A self-consistent model of an isothermal tokamak
McNamara, Steven; Lilley, Matthew
2014-10-01
Continued progress in liquid lithium coating technologies have made the development of a beam driven tokamak with minimal edge recycling a feasibly possibility. Such devices are characterised by improved confinement due to their inherent stability and the suppression of thermal conduction. Particle and energy confinement become intrinsically linked and the plasma thermal energy content is governed by the injected beam. A self-consistent model of a purely beam fuelled isothermal tokamak is presented, including calculations of the density profile, bulk species temperature ratios and the fusion output. Stability considerations constrain the operating parameters and regions of stable operation are identified and their suitability to potential reactor applications discussed.
Geodesic acoustic modes in noncircular cross section tokamaks
Energy Technology Data Exchange (ETDEWEB)
Sorokina, E. A., E-mail: sorokina.ekaterina@gmail.com; Lakhin, V. P. [National Research Center “Kurchatov Institute,” (Russian Federation); Konovaltseva, L. V. [People’s Friendship University of Russia (Russian Federation); Ilgisonis, V. I. [National Research Center “Kurchatov Institute,” (Russian Federation)
2017-03-15
The influence of the shape of the plasma cross section on the continuous spectrum of geodesic acoustic modes (GAMs) in a tokamak is analyzed in the framework of the MHD model. An expression for the frequency of a local GAM for a model noncircular cross section plasma equilibrium is derived. Amendments to the oscillation frequency due to the plasma elongation and triangularity and finite tokamak aspect ratio are calculated. It is shown that the main factor affecting the GAM spectrum is the plasma elongation, resulting in a significant decrease in the mode frequency.
Numerical simulation for HT-6M tokamak electrical transient behaviours
International Nuclear Information System (INIS)
Yu Yuanqi; Liu Baohua; Pan Yuan
1991-02-01
The following main points are concerned: (1) State equations used for dynamic analysis of all electrical parameters of the tokamak are derived. (2) In order to increase plasma volt-seconds and to get plasma current with longer sustainment phase, a power supply scheme for HT-6M and its numerical simulation are studied. (3) The distribution of energy flow in coupling loops of the tokamak is discussed, and the energy transfer ratio from the OH loop and vertical field loop to the plasma is also analyzed
Economic trends of tokamak power plants independent of physics scaling models
International Nuclear Information System (INIS)
Reid, R.L.; Steiner, D.
1978-01-01
This study examines the effects of plasma radius, field on axis, plasma impurity level, and aspect ratio on power level and unit capital cost, $/kW/sub e/, of tokamak power plants sized independent of plasma physics scaling models. It is noted that tokamaks sized in this manner are thermally unstable based on trapped particle scaling relationships. It is observed that there is an economic advantage for larger power level tokamaks achieved by physics independent sizing; however, the incentive for increased power levels is less than that for fission reactors. It is further observed that the economic advantage of these larger power level tokamaks is decreased when plasma thermal stability measures are incorporated, such as by increasing the plasma impurity concentration. This trend of economy with size obtained by physics independent sizing is opposite to that observed when the tokamak designs are constrained to obey the trapped particle and empirical scaling relationships
International Nuclear Information System (INIS)
Rome, J.A.; Peng, Y.K.M.
1978-09-01
Guiding center orbits in noncircular axisymmetric tokamak plasmas are studied in the constants of motion (COM) space of (v, zeta, psi/sub m/). Here, v is the particle speed, zeta is the pitch angle with respect to the parallel equilibrium current, J/sub parallels/, and psi/sub m/ is the maximum value of the poloidal flux function (increasing from the magnetic axis) along the guiding center orbit. Two D-shaped equilibria in a flux-conserving tokamak having β's of 1.3% and 7.7% are used as examples. In this space, each confined orbit corresponds to one and only one point and different types of orbits (e.g., circulating, trapped, stagnation and pinch orbits) are represented by separate regions or surfaces in the space. It is also shown that the existence of an absolute minimum B in the higher β (7.7%) equilibrium results in a dramatically different orbit topology from that of the lower β case. The differences indicate the confinement of additional high energy (v → c, within the guiding center approximation) trapped, co- and countercirculating particles whose orbit psi/sub m/ falls within the absolute B well
International Nuclear Information System (INIS)
Doggett, J.; Salpietro, E.; Shatalov, G.
1991-01-01
The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs
International Nuclear Information System (INIS)
Rudakov, D; Yu, J; Boedo, J; Hollmann, E; Krasheninnikov, S; Moyer, R; Muller, S; Yu, A; Rosenberg, M; Smirnov, R; West, W; Boivin, R; Bray, B; Brooks, N; Hyatt, A; Wong, C; Fenstermacher, M; Groth, M; Lasnier, C; McLean, A; Stangeby, P; Ratynskaia, S; Roquemore, A; Skinner, C; Solomon, W M
2008-01-01
Dust production and accumulation impose safety and operational concerns for ITER. Diagnostics to monitor dust levels in the plasma as well as in-vessel dust inventory are currently being tested in a few tokamaks. Dust accumulation in ITER is likely to occur in hidden areas, e.g. between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering resolves size of particles between 0.16-1.6 (micro)m in diameter; the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast-framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in 2D with a single camera or 3D using multiple cameras, but determination of particle size is problematic. In order to calibrate diagnostics and benchmark dust dynamics modeling, pre-characterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase of carbon atomic, C2 dimer, and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics
Axisymmetric control in tokamaks
International Nuclear Information System (INIS)
Humphreys, D.A.
1991-02-01
Vertically elongated tokamak plasmas are intrinsically susceptible to vertical axisymmetric instabilities as a result of the quadrupole field which must be applied to produce the elongation. The present work analyzes the axisymmetric control necessary to stabilize elongated equilibria, with special application to the Alcator C-MOD tokamak. A rigid current-conserving filamentary plasma model is applied to Alcator C-MOD stability analysis, and limitations of the model are addressed. A more physically accurate nonrigid plasma model is developed using a perturbed equilibrium approach to estimate linearized plasma response to conductor current variations. This model includes novel flux conservation and vacuum vessel stabilization effects. It is found that the nonrigid model predicts significantly higher growth rates than predicted by the rigid model applied to the same equilibria. The nonrigid model is then applied to active control system design. Multivariable pole placement techniques are used to determine performance optimized control laws. Formalisms are developed for implementing and improving nominal feedback laws using the C-MOD digital-analog hybrid control system architecture. A proportional-derivative output observer which does not require solution of the nonlinear Ricatti equation is developed to help accomplish this implementation. The nonrigid flux conserving perturbed equilibrium plasma model indicates that equilibria with separatrix elongation of at least κ sep = 1.85 can be stabilized robustly with the present control architecture and conductor/sensor configuration
Modelling dust transport in tokamaks
International Nuclear Information System (INIS)
Martin, J.D.; Martin, J.D.; Bacharis, M.; Coppins, M.; Counsell, G.F.; Allen, J.E.; Counsell, G.F.
2008-01-01
The DTOKS code, which models dust transport through tokamak plasmas, is described. The floating potential and charge of a dust grain in a plasma and the fluxes of energy to and from it are calculated. From this model, the temperature of the dust grain can be estimated. A plasma background is supplied by a standard tokamak edge modelling code (B2SOLPS5.0), and dust transport through MAST (the Mega-Amp Spherical Tokamak) and ITER plasmas is presented. We conclude that micron-radius tungsten dust can reach the separatrix in ITER. (authors)
Deposit of thin films for Tokamaks conditioning
International Nuclear Information System (INIS)
Valencia A, R.
2006-01-01
discharge plasma, created in a calibrated mixture of methane-hydrogen during the hydrogenated amorphous carbon film deposit on the vessel wall of Novillo tokamak, were determined by mass spectrometry. By way of measuring the emission lines of the carbon and oxygen impurities in intense discharges, the time required by the plasma to interact with the wall was estimated. In addition to it, the temporal conduct of the emission line intensity of these impurities was observed by means of an intensified CCD detector. Once an ∼ 10 % of helium was introduced in the operating gas of the tokamak discharges, a 25-42 eV time variation of the electron temperature was measured using the intensity ratio technique. (Author)
Tokamak building-design considerations for a large tokamak device
International Nuclear Information System (INIS)
Barrett, R.J.; Thomson, S.L.
1981-01-01
Design and construction of a satisfactory tokamak building to support FED appears feasible. Further, a pressure vessel building does not appear necessary to meet the plant safety requirements. Some of the building functions will require safety class systems to assure reliable and safe operation. A rectangular tokamak building has been selected for FED preconceptual design which will be part of the confinement system relying on ventilation and other design features to reduce the consequences and probability of radioactivity release
Natural current profiles in tokamaks
International Nuclear Information System (INIS)
Biskamp, D.
1986-01-01
It is proposed that a certain class of equilibrium, which follow from an elementary variational principle, are the natural current profiles in tokamaks, to which actual discharge profiles tend to relax. (orig.)
Federal Laboratory Consortium — Alcator C-Mod at the Massachusetts Institute of Technology is operated as a DOE national user facility. Alcator C-Mod is a unique, compact tokamak facility that uses...
JUST: Joint Upgraded Spherical Tokamak
International Nuclear Information System (INIS)
Azizov, E.A.; Dvorkin, N.Ya.; Filatov, O.G.
1997-01-01
The main goals, ideas and the programme of JUST, spherical tokamak (ST) for the plasma burn investigation, are presented. The place and prospects of JUST in thermonuclear investigations are discussed. (author)
Preliminary Design of Alborz Tokamak
Mardani, M.; Amrollahi, R.; Saramad, S.
2012-04-01
The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. The most important part of the tokamak design is the design of TF coils. In this paper a refined design of the TF coil system for the Alborz tokamak is presented. This design is based on cooper cable conductor with 5 cm width and 6 mm thickness. The TF coil system is consist of 16 rectangular shape coils, that makes the magnetic field of 0.7 T at the plasma center. The stored energy in total is 160 kJ, and the power supply used in this system is a capacitor bank with capacity of C = 1.32 mF and V max = 14 kV.
International Nuclear Information System (INIS)
Varandas, C.A.F.; Cabral, J.A.C.; Manso, M.E.
1991-01-01
A small tokamak is under construction at the Portuguese Technical Superior Institute. The main objective is to create a home based laboratory in which an independent scientific program might be developed. (L.C.J.A.). 14 refs, 6 figs
Numerical Tokamak Project code comparison
International Nuclear Information System (INIS)
Waltz, R.E.; Cohen, B.I.; Beer, M.A.
1994-01-01
The Numerical Tokamak Project undertook a code comparison using a set of TFTR tokamak parameters. Local radial annulus codes of both gyrokinetic and gyrofluid types were compared for both slab and toroidal case limits assuming ion temperature gradient mode turbulence in a pure plasma with adiabatic electrons. The heat diffusivities were found to be in good internal agreement within ± 50% of the group average over five codes
Spherical tokamak development in Brazil
International Nuclear Information System (INIS)
Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J.; Barbosa, L.F.W.; Patire Junior, H.; The high-power microwave sources group
2003-01-01
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Confinement and diffusion in tokamaks
International Nuclear Information System (INIS)
McWilliams, R.
1988-01-01
The effect of electric field fluctuations on confinement and diffusion in tokamak is discussed. Based on the experimentally determined cross-field turbolent diffusion coefficient, D∼3.7*cT e /eB(δn i /n i ) rms which is also derived by a simple theory, the cross-field diffusion time, tp=a 2 /D, is calculated and compared to experimental results from 51 tokamak for standard Ohmic operation
Enhancement of confinement in tokamaks
International Nuclear Information System (INIS)
Furth, H.P.
1986-01-01
The analysis begins by identifying a hypothetical model of tokamak confinement that is designed to take into account the conflict between Tsub(e)(r)-profile shapes arising from microscopic transport and J(r)-profile shapes required for gross stability. On the basis of this model, a number of hypothetical lines of advance are developed. Some TFTR experiments that may point the way to a particularly attractive type of tokamak reactor regime are discussed. (author)
Spherical tokamak development in Brazil
Energy Technology Data Exchange (ETDEWEB)
Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma] (and others)
2003-07-01
The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Spherical tokamak development in Brazil
Energy Technology Data Exchange (ETDEWEB)
Ludwig, G.O.; Del Bosco, E.; Ferreira, J.G.; Berni, L.A.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Ueda, M.; Barroso, J.J.; Castro, P.J. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma; Barbosa, L.F.W. [Universidade do Vale do Paraiba (UNIVAP), Sao Jose dos Campos, SP (Brazil). Faculdade de Engenharia, Arquitetura e Urbanismo; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Div. de Mecanica Espacial e Controle; The high-power microwave sources group
2003-12-01
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the steps in the development of the ETE (Experimento Tokamak Esferico) project, its research program, technical characteristics and operating conditions as of December, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Spherical tokamak development in Brazil
International Nuclear Information System (INIS)
Ludwig, Gerson Otto; Bosco, Edson Del; Ferreira, Julio Guimaraes
2003-01-01
The general characteristics of spherical tokamaks, or spherical tori, with a brief view of work in this area already performed or in progress at several institutions worldwide are described. The paper presents also the steps in the development of the ETE (Experiment Tokamak spheric) project, its research program, technical characteristics and operating conditions as of December, 2002 a the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
International Nuclear Information System (INIS)
Navratil, G.A.; Mauel, M.E.; Ivers, T.H.; Sankar, M.K.V.; Eisner, E.; Gates, D.; Garofalo, A.; Kombargi, R.; Maurer, D.; Nadle, D.; Xiao, Q.
1993-01-01
During the past 6 months, experiments have been conducted with the HBT-EP tokamak in order to (1) test and evaluate diagnostic systems, (2) establish basic machine operation, (3) document MHD behavior as a function of global discharge parameters, (4) investigate conditions leading to passive stabilization of MHD instabilities, and (5) quantify the external saddle coil current required for DC mode locking. In addition, the development and installation of new hardware systems has occurred. A prototype saddle coil was installed and tested. A five-position (n,m) = (1,2) external helical saddle coil was attached for mode-locking experiments. And, fabrication of the 32-channel UV tomography and the multipass Thomson scattering diagnostics have begun in preparation for installation later this year
Anomalous transport in tokamaks
International Nuclear Information System (INIS)
Wootton, A.J.
1989-01-01
A review is presented of what is known about anomalous transport in tokamaks. It is generally thought that this anomalous transport is the result of fluctuations in various plasma parameters. In the plasma edge detailed measurements of the quantities required to directly determine the fluctuation driven fluxes are available. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge energy flux are less clear, but electrostatic convection plays an important part. In the confinement region experimental observations are presently restricted to measurements of density and potential fluctuations and their correlations. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made. Magnetic fluctuations is discussed
International Nuclear Information System (INIS)
Tenney, F.H.
1976-09-01
A report on one year of study of a tokamak hybrid reactor is presented. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters
International Nuclear Information System (INIS)
Tenney, F.H.
1976-01-01
A report on one year of study of a tokamak hybrid reactor is given. The plasma is maintained by both D and T beams. To obtain long burn times a poloidal field divertor is required. Both the single null and the double null style of divertor are considered. The blanket consists of a neutron multiplier region containing natural uranium followed by burner regions of molten salt (flibe) loaded with PuF 3 to enhance the energy multiplication. Economic analysis has been applied only recently to a variety of reactor sizes and plasma conditions. Early indications suggest that the most attractive hybrids will have large plasmas of major radius in excess of 8 meters
International Nuclear Information System (INIS)
Schmidt, J.
1987-01-01
The author discusses his lab's plan for completing the Compact Ignition Tokamak (CIT) conceptual design during calendar year 1987. Around July 1 they froze the subsystem envelopes on the device to continue with the conceptual design. They did this by formalizing a general requirements document. They have been developing the management plan and submitted a version to the DOE July 10. He describes a group of management activities. They released the vacuum vessel Request For Proposals (RFP) on August 5. An RFP to do a major part of the system engineering on the device is being developed. They intend to assemble the device outside of the test cell, then move it into the the test cell, install it there, and bring to the test cell many of the auxiliary facilities from TFTR, for example, power supplies
Energy Technology Data Exchange (ETDEWEB)
Caldas, Ibere L.; Heller, M.V.A.P.; Brasilio, Z.A. [Sao Paulo Univ., SP, RJ (Brazil). Inst. de Fisica
1997-12-31
Full text. In this work we summarize the results from experiments on electrostatic and magnetic fluctuations in tokamak plasmas. Spectral analyses show that these fluctuations are turbulent, having a broad spectrum of wavectors and a broad spectrum of frequencies at each wavector. The electrostatic turbulence induces unexpected anomalous particle transport that deteriorates the plasma confinement. The relationship of these fluctuations to the current state of plasma theory is still unclear. Furthermore, we describe also attempts to control this plasma turbulence with external magnetic perturbations that create chaotic magnetic configurations. Accordingly, the magnetic field lines may become chaotic and then induce a Lagrangian diffusion. Moreover, to discuss nonlinear coupling and intermittency, we present results obtained by using numerical techniques as bi spectral and wavelet analyses. (author)
International Nuclear Information System (INIS)
Bondeson, A.
1987-01-01
This paper discusses major and minor disruptions in Tokamaks. A number of models and numerical simulations of disruptions based on resistive MHD are reviewed. A discussion is given of how disruptive current profiles are correlated with the experimentally known operational limits in density and current. It is argued that the q a =2 limit is connected with stabilization of the m=2/n=1 tearing mode for a approx.< 2.7 by resistive walls and mode rotation. Experimental and theoretical observations indicate that major disruptions usually occur in at least two phases, first a 'predisruption', or loss of confinement in the region 1 < q < 2, leaving the q approx.= 1 region almost unaffected, followed by a final disruption of the central part, interpreted here as a toroidal n = 1 external kink mode. (author)
The ARIES tokamak reactor study
International Nuclear Information System (INIS)
1989-10-01
The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions
HTMR: an experimental tokamak reactor with hybrid copper/superconductor toroidal field magnet
International Nuclear Information System (INIS)
Avanzini, P.G.; Raia, G.; Rosatelli, F.; Zampaglione, V.
1985-01-01
The feasibility of a hybrid configuration superconducting coils/copper coils for a next generation tokamak TF magnet has been investigated. On the basis of this hybrid solution, the conceptual design has been developed for a medium-high toroidal field tokamak reactor (HTMR). The results of this study show the possibility of designing a tokamak reactor with reduced size in comparison with other INTOR like devices, still gaining some margins in front of the uncertainties in the scaling laws for plasma physics parameters and retaining the presence of a blanket with a tritium breeding ratio of about 1
Energy Technology Data Exchange (ETDEWEB)
Barbosa, Luis Filipe F.P.W.; Bosco, Edson del
1994-12-31
This report presents the project and analysis of the circuit for production of the toroidal magnetic field in the Tokamak ETE (Spherical Tokamak Experiment). The ETE is a Tokamak with a small-aspect-ratio parameter to be used for studying the plasma physics for the research on thermonuclear fusion. This machine is being constructed at the Laboratorio Associado de Plasma (LAP) of the Instituto Nacional de Pesquisas Espaciais (INPE) in Sao Jose dos Campos, SP, Brazil. (author). 20 refs., 39 figs., 4 tabs.
Reusch, Joshua
2017-10-01
A major goal of the spherical tokamak research program is accessing a state of low internal inductance li, high elongation κ, high toroidal and normalized beta (βt and βN) , and low collisionality without solenoidal current drive. A new local helicity injection (LHI) system in the lower divertor region of the ultra-low aspect ratio Pegasus ST provides non-solenoidally driven plasmas that exhibit most of these characteristics. LHI utilizes compact, edge-localized current sources (Ainj 4 cm2, Iinj 8 kA, Vinj 1.5 kV) for plasma startup and sustainment, and can sustain more than 200 kA of plasma current. Plasma growth via LHI is enhanced by a transition from a regime of high kink-like MHD activity to one of reduced MHD activity at higher frequencies and presumably shorter wavelengths. The strong edge current drive provided by LHI results in a hollow current density profile with low li. The low aspect ratio (R0 / a 1.2) of Pegasus allows ready access to high κ and MHD stable operation at very high normalized plasma currents (IN =Ip /aBT> 15). Thomson scattering measurements indicate Te 100 eV and ne 1 ×19 m-3. The impurity Ti evolution is correlated in time with high frequency magnetic fluctuations, implying substantial reconnection ion heating is driven by the applied helicity injection. Doppler spectroscopy indicates Ti >=Te and that the anomalous ion heating scales consistently with two fluid reconnection theory. Taken together, these features provide access to very high βt plasmas. Equilibrium analyses indicate βt up to 100% and βN 6.5 is achieved. At increasingly low BT, the discharge disrupts at the no-wall ideal stability limit. In these high βt discharges, a minimum |B| well forms over 50% of the plasma volume. This unique magnetic configuration may be of interest for testing predictions of stabilizing drift wave turbulence and/or improving energetic particle confinement. This work supported by US DOE Grants DE-FG02-96ER54375 and DE-SC0006928.
The physics of an ignited tokamak
International Nuclear Information System (INIS)
Troyon, F.
1990-10-01
There appears to be a consensus that time has come to embark on the design and construction of the next generation of tokamaks which is at the origin of the ITER initiative. Different proposals have been made based on different appreciation as to the size of the step which can be taken, related to considerations of cost, risk and duration of construction. A class of devices which may be considered the last the very high-field, high density ALCATOR-Frascati line of tokamaks have been proposed for some years specifically for this purpose. Today there remain three such projects: Ignitor, Ignitex and CIT. The technology chosen limits the pulse length to a few seconds. These devices have evolved through the years becoming larger and much more expensive than originally anticipated, increasing the pressure to do more than just a simple demonstration of ignition. There is another class of more ambitious devices which aim at creating long burning plasmas in conditions as close as possible to those of a tokamak reactor in order to address all the plasma physics problems associated with long burn. Three such projects, NET, the european next step after JET, ITER and JIT are good examples of this approach. The ideal would be to design a device with sufficient margin to study burning plasmas over a wide range of parameters. The object of this didactic presentation is to describe the common physics basis of all these projects, compare their expected performance using present knowledge and list the physics problems associated with a burning plasma experiment. The comparison is not meant to be a judgement since the important parameter is the cost/benefit ratio which is a matter of appreciation at this stage. 6 refs., 3 figs., 1 tab
Saturated tearing modes in tokamaks with divertors
International Nuclear Information System (INIS)
Bateman, G.
1982-12-01
We have developed a self-consistent theory of saturated tearing modes capable of predicting multiple magnetic island widths in tokamaks with no assumptions on the cross-sectional shape, aspect ratio, or plasma pressure. We are in the process of implementing this algorithm in the form of a computer code. We propose: (1) to complete, refine, document and publish this computer code; (2) to carry out a survey in which we vary the current profile, aspect ratio, cross-sectional shape, and pressure profile in order to determine their effect on saturated tearing mode magnetic island widths; and (3) to determine the effect of some externally applied magnetic perturbation harmonics on these magnetic island widths. Particular attention will be paid to the coupling between different helical harmonics, the effect of multiple magnetic islands on the profiles of temperature, pressure and current, and the potential of magnetic island overlap leading to a disruptive instability
Bibliography of fusion product physics in tokamaks
International Nuclear Information System (INIS)
Hively, L.M.; Sigmar, D.J.
1989-09-01
Almost 700 citations have been compiled as the first step in reviewing the recent research on tokamak fusion product effects in tokamaks. The publications are listed alphabetically by the last name of the first author and by subject category
Digital controlled pulsed electric system of the ETE tokamak. First report
International Nuclear Information System (INIS)
Barbosa, Luis Felipe de F.P.W.; Del Bosco, Edson
1997-01-01
This reports presents a summary on the thermonuclear fusion and application for energy supply purposes. The tokamak device operation and the magnetic field production systems are described. The ETE tokamak is a small aspect ratio device designed for plasma physics and thermonuclear fusion studies, which presently is under construction at the Laboratorio Associado de Plasma (LAP), Instituto Nacional de Pesquisas Espaciais (INPE) - S.J. dos Campos - S. Paulo. (author)
Energy, Vacuum, Gas Fueling, and Security Systems for the Spherical Tokamak MEDUSA-CR
Gonzalez, Jeferson; Soto, Christian; Carvajal, Johan; Ribeiro, Celso
2013-10-01
The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R security systems for MEDUSA-CR device. The interface with the control and data acquisition systems based on National Instruments (NI) software (LabView) and hardware (on loan to our laboratory via NI-Costa Rica) are also presented. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.
Moving Divertor Plates in a Tokamak
International Nuclear Information System (INIS)
Zweben, S.J.; Zhang, H.
2009-01-01
Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions
Fusion potential for spherical and compact tokamaks
International Nuclear Information System (INIS)
Sandzelius, Mikael
2003-02-01
The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high β-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect
Fusion potential for spherical and compact tokamaks
Energy Technology Data Exchange (ETDEWEB)
Sandzelius, Mikael
2003-02-01
The tokamak is the most successful fusion experiment today. Despite this, the conventional tokamak has a long way to go before being realized into an economically viable power plant. In this master thesis work, two alternative tokamak configurations to the conventional tokamak has been studied, both of which could be realized to a lower cost. The fusion potential of the spherical and the compact tokamak have been examined with a comparison of the conventional tokamak in mind. The difficulties arising in the two configurations have been treated from a physical point of view concerning the fusion plasma and from a technological standpoint evolving around design, materials and engineering. Both advantages and drawbacks of either configuration have been treated relative to the conventional tokamak. The spherical tokamak shows promising plasma characteristics, notably a high {beta}-value but have troubles with high heat loads and marginal tritium breeding. The compact tokamak operates at a high plasma density and a high magnetic field enabling it to be built considerably smaller than any other tokamak. The most notable down-side being high heat loads and neutron transport problems. With the help of theoretical reactor studies, extrapolating from where we stand today, it is conceivable that the spherical tokamak is closer of being realized of the two. But, as this study shows, the compact tokamak power plant concept offers the most appealing prospect.
Moving Divertor Plates in a Tokamak
Energy Technology Data Exchange (ETDEWEB)
S.J. Zweben, H. Zhang
2009-02-12
Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.
Tokamak Physics Experiment (TPX) design
International Nuclear Information System (INIS)
Schmidt, J.A.
1995-01-01
TPX is a national project involving a large number of US fusion laboratories, universities, and industries. The element of the TPX requirements that is a primary driver for the hardware design is the fact that TPX tokamak hardware is being designed to accommodate steady state operation if the external systems are upgraded from the 1,000 second initial operation. TPX not only incorporates new physics, but also pioneers new technologies to be used in ITER and other future reactors. TPX will be the first tokamak with fully superconducting magnetic field coils using advanced conductors, will have internal nuclear shielding, will use robotics for machine maintenance, and will remove the continuous, concentrated heat flow from the plasma with new dispersal techniques and with special materials that are actively cooled. The Conceptual Design for TPX was completed during Fiscal Year 1993. The Preliminary Design formally began at the beginning of Fiscal Year 1994. Industrial contracts have been awarded for the design, with options for fabrication, of the primary tokamak hardware. A large fraction of the design and R and D effort during FY94 was focused on the tokamak and in turn on the tokamak magnets. The reason for this emphasis is because the magnets require a large design and R and D effort, and are critical to the project schedule. The magnet development is focused on conductor development, quench protection, and manufacturing R and D. The Preliminary Design Review for the Magnets is planned for fall, 1995
Resistive instabilities in tokamaks
International Nuclear Information System (INIS)
Rutherford, P.H.
1985-10-01
Low-m tearing modes constitute the dominant instability problem in present-day tokamaks. In this lecture, the stability criteria for representative current profiles with q(0)-values slightly less than unit are reviewed; ''sawtooth'' reconnection to q(0)-values just at, or slightly exceeding, unity is generally destabilizing to the m = 2, n = 1 and m = 3, n = 2 modes, and severely limits the range of stable profile shapes. Feedback stabilization of m greater than or equal to 2 modes by rf heating or current drive, applied locally at the magnetic islands, appears feasible; feedback by island current drive is much more efficient, in terms of the radio-frequency power required, then feedback by island heating. Feedback stabilization of the m = 1 mode - although yielding particularly beneficial effects for resistive-tearing and high-beta stability by allowing q(0)-values substantially below unity - is more problematical, unless the m = 1 ideal-MHD mode can be made positively stable by strong triangular shaping of the central flux surfaces. Feedback techniques require a detectable, rotating MHD-like signal; the slowing of mode rotation - or the excitation of non-rotating modes - by an imperfectly conducting wall is also discussed
Tokamak experimental power reactor
International Nuclear Information System (INIS)
Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.
1978-01-01
A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars
Classical tokamak transport theory
International Nuclear Information System (INIS)
Nocentini, Aldo
1982-01-01
A qualitative treatment of the classical transport theory of a magnetically confined, toroidal, axisymmetric, two-species plasma is presented. The 'weakly collisional' ('banana' and 'plateau') and 'collision dominated' ('Pfirsch-Schlueter' and 'highly collisional') regimes, as well as the Ware effect are discussed. The method used to evaluate the diffusion coffieicnts of particles and heat in the weakly collisional regime is based on stochastic argument, that requires an analysis of the characteristic collision frequencies and lengths for particles moving in a tokamak-like magnetic field. The same method is used to evaluate the Ware effect. In the collision dominated regime on the other hand, the particle and heat fluxes across the magnetic field lines are dominated by macroscopic effects so that, although it is possible to present them as diffusion (in fact, the fluxes turn out to be proportional to the density and temperature gradients), a macroscopic treatment is more appropriate. Hence, fluid equations are used to inveatigate the collision dominated regime, to which particular attention is devoted, having been shown relatively recently that it is more complicated than the usual Pfirsch-Schlueter regime. The whole analysis presented here is qualitative, aiming to point out the relevant physical mechanisms involved in the various regimes more than to develop a rigorous mathematical derivation of the diffusion coefficients, for which appropriate references are given. (author)
Tokamak experimental power reactor
International Nuclear Information System (INIS)
Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.
1976-01-01
A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors
Numerical study for determining PF coil system parameters in MHD equilibrium of KT-2 tokamak plasma
International Nuclear Information System (INIS)
Ryu, J.; Hong, S.H.; Lee, K.W.; Hong, B.G.; In, S.R.; Kim, S.K.
1995-01-01
The KT-2 is a large-aspect-ratio medium-sized divertor tokamak in the conceptual design phase and planned to be operational in 1998 at the Korea Atomic Energy Research Institute (KAERI). Plasma equilibrium in tokamak can be acquired by controlling the current of poloidal field (PF) coils in appropriate geometries and positions. In this study, the authors have performed numerical calculations to achieve the various equilibrium conditions fitting given plasma shapes and satisfying PF current limitations. Usually an ideal magnetohydrodynamic (MHD) equation is used to obtain the equilibrium solution of tokamak plasma, and it is practical to take advantage of a numerical method in solving the MHD equation because it has nonlinear source terms. Two equilibrium codes have been applied to find a double-null configuration of free-boundary tokamak plasma in KT-2: one is of the authors' own developing and the other is a free-boundary tokamak equilibrium code (FBT) that has been used mainly for the verification of developed code's results. PF coil system parameters including their positions and currents are determined for the optimization of input power required when the specifications of KT-2 tokamak are met. Then, several sets of equilibrium conditions during the tokamak operation are found to observe the changes of poloidal field currents with the passing of operation time step, and the basic stability problems related with the magnetic field structure is also considered
Performance and stability limits at near-unity aspect ratio in the pegasus toroidal experiment
International Nuclear Information System (INIS)
Fonck, R.; Diem, S.; Garstka, G.; Kissick, M.; Lewicki, B.; Ostrander, C.; Probert, P.; Reinke, M.; Sontag, A.; Tritz, K.; Unterberg, E.
2003-01-01
The Pegasus Toroidal Experiment is a mid-sized extremely-low aspect ratio (A) spherical torus (ST). It has the dual roles of exploring limits of ST behavior as A approaches 1 and studying the physics of ST plasmas in the tokamak-spheromak overlap regime. Major parameters are R 0.25 - 0.45 m, A 1.1 - 1.4, I p T 20% have been obtained, and the operational space of beta vs I p /aB T is similar to that observed for NBI-heated START discharges. Achievable plasma current is subject to an apparent limit of I p /I tf ∼ 1. Access to higher-current plasmas appears to be restricted by the appearance of large internal MHD activity, including m/n=2/1 and 3/2 modes. Recent experiments have begun to access ideal stability limits, with disruptions observed as q 95 approaches 5, in agreement with numerical predictions for external kink mode onset. (author)
Dynamic stabilization of disruption precursors in tokamak
Energy Technology Data Exchange (ETDEWEB)
Maoquan, Wang; Jianshan, Mao; Yuan, Pan [Academia Sinica, Hefei, AH (China). Inst. of Plasma Physics
1994-12-01
A method for dynamic stabilization of the disruption precursors in tokamak is proposed, that is a controlled ac current induced and added to the equilibrium current. The ac currents applied can be a sine alternative current with a relevant frequency, or a pulsed current with a suitable pulsed width {tau} and or a discontinuous pulsed current whose width {tau} is very shorter than the intervals between pulses, and or a `sawtooth` pulsed current with the time of ramp phase of the sawtooth is very much shorter than the sawtooth descending time, the ratio of them can be {<=}10{sup -3}. The physical model of the ac current drive is analyzed in detail. The suppression role of the ac current on the MHD perturbations was analyzed in theory and proved numerically. It is indicated that the ac current can make the discontinuous derivative, {Delta}`, more favorable for the tearing mode stabilities, and so, as long as the parameters of the applied ac currents are selected suitably, the MHD perturbations can be suppressed effectively, the perturbations will be in the zero-growing state, the profile of the plasma current and temperature remain in the initial states and not variate basically, the tokamak be in the stabilized operation state. (8 figs.).
Tokamak power systems studies at ANL
International Nuclear Information System (INIS)
Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.
1986-01-01
A number of advances in plasma physics and engineering promise to greatly improve the reactor prospects of tokamaks. The following features, in particular, are examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 7 T; (c) low toroidal current (I ≅ 5MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields produced in the plasma. In addition to matching desirable high-beta equilibria, this method is capable of producing a large variety of new equilibria, many of which look attractive. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors
Advanced tokamak burning plasma experiment
International Nuclear Information System (INIS)
Porkolab, M.; Bonoli, P.T.; Ramos, J.; Schultz, J.; Nevins, W.N.
2001-01-01
A new reduced size ITER-RC superconducting tokamak concept is proposed with the goals of studying burn physics either in an inductively driven standard tokamak (ST) mode of operation, or in a quasi-steady state advanced tokamak (AT) mode sustained by non-inductive means. This is achieved by reducing the radiation shield thickness protecting the superconducting magnet by 0.34 m relative to ITER and limiting the burn mode of operation to pulse lengths as allowed by the TF coil warming up to the current sharing temperature. High gain (Q≅10) burn physics studies in a reversed shear equilibrium, sustained by RF and NB current drive techniques, may be obtained. (author)
Plasma boundary phenomena in tokamaks
International Nuclear Information System (INIS)
Stangeby, P.C.
1989-06-01
The focus of this review is on processes occurring at the edge, and on the connection between boundary plasma - the scrape-off layer (SOL) and the radiating layer - and central plasma processes. Techniques used for edge diagnosis are reviewed and basic experimental information (n e and T e ) is summarized. Simple models of the SOL are summarized, and the most important effects of the boundary plasma - the influence on the fuel particles, impurities, and energy - on tokamak operation dealt with. Methods of manipulating and controlling edge conditions in tokamaks and the experimental data base for the edge during auxiliary heating of tokamaks are reviewed. Fluctuations and asymmetries at the edge are also covered. (9 tabs., 134 figs., 879 refs.)
Computational studies of tokamak plasmas
International Nuclear Information System (INIS)
Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji
1981-02-01
Computational studies of tokamak plasmas are extensively advanced. Many computational codes have been developed by using several kinds of models, i.e., the finite element formulation of MHD equations, the time dependent multidimensional fluid model, and the particle model with the Monte-Carlo method. These codes are applied to the analyses of the equilibrium of an axisymmetric toroidal plasma (SELENE), the time evolution of the high-beta tokamak plasma (APOLLO), the low-n MHD stability (ERATO-J) and high-n ballooning mode stability (BOREAS) in the INTOR tokamak, the nonlinear MHD stability, such as the positional instability (AEOLUS-P), resistive internal mode (AEOLUS-I) etc., and the divertor functions. (author)
Development in Diagnostics Application to Control Advanced Tokamak Plasma
International Nuclear Information System (INIS)
Koide, Y.
2008-01-01
For continuous operation expected in DEMO, all the plasma current must be non-inductively driven, with self-generated neoclassical bootstrap current being maximized. The control of such steady state high performance tokamak plasma (so-called 'Advanced Tokamak Plasma') is a challenge because of the strong coupling between the current density, the pressure profile and MHD stability. In considering diagnostic needs for the advanced tokamak research, diagnostics for MHD are the most fundamental, since discharges which violate the MHD stability criteria either disrupt or have significantly reduced confinement. This report deals with the development in diagnostic application to control advanced tokamak plasma, with emphasized on recent progress in active feedback control of the current profile and the pressure profile under DEMO-relevant high bootstrap-current fraction. In addition, issues in application of the present-day actuators and diagnostics for the advanced control to DEMO will be briefly addressed, where port space for the advanced control may be limited so as to keep sufficient tritium breeding ratio (TBR)
Study of a compact reversed shear Tokamak reactor
International Nuclear Information System (INIS)
Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.
1998-01-01
A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)
Core fuelling to produce peaked density profiles in large tokamaks
International Nuclear Information System (INIS)
Mikkelsen, D.R.; McGuire, K.M.; Schmidt, G.L.; Zweben, S.J.
1995-01-01
Peaking the density profile increases the usable bootstrap current and the average fusion power density; this could reduce the current drive power and increase the net output of power producing tokamaks. The use of neutral beams and pellet injection to produce peaked density profiles is assessed. It is shown that with radially 'hollow' diffusivity profiles (and no particle pinch) moderately peaked density profiles can be produced by particle source profiles that are peaked off-axis. The fuelling penetration requirements can therefore be relaxed and this greatly improves the feasibility of generating peaked density profiles in large tokamaks. In particular, neutral beam fuelling does not require Megavolt particle energies. Even with beam voltages of ∼ 200 keV, however, exceptionally good particle confinement is needed to achieve net electrical power generation. The required ratio of particle to thermal diffusivities is an order of magnitude outside the range reported for tokamaks. In a system with no power production requirement (e.g., neutron sources) neutral beam fuelling should be capable of producing peaked density profiles in devices as large as ITER. Fuelling systems with low energy cost per particle - such as cryogenic pellet injection - must be used in power producing tokamaks when τ P ∼ τ E . Simulations with pellet injection speeds of 7 km/s show that the peaking factor, n e0 / e >, approaches 2. (author). 65 refs, 8 figs
Summary discussion: An integrated advanced tokamak reactor
International Nuclear Information System (INIS)
Sauthoff, N.R.
1994-01-01
The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ''figures of merit'' for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept
STARFIRE: a commercial tokamak reactor
International Nuclear Information System (INIS)
1979-12-01
The purpose of this document is to provide an interim status report on the STARFIRE project for the period of May to September 1979. The basic objective of the STARFIRE project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor
LHCD experiments on tokamak CASTOR
International Nuclear Information System (INIS)
Zacek, F.; Badalec, J.; Jakubka, J.; Kryska, L.; Preinhaelter, J.; Stoeckel, J.; Valovic, M.; Nanobashvili, S.; Weixelbaum, L.; Wenzel, U.; Spineanu, F.; Vlad, M.
1990-10-01
A short survey is given of the experimental activities at the small Prague tokamak CASTOR. They concern primarily the LH current drive using multijunction waveguide grills as launching antennae. During two last years the, efforts were focused on a study of the electrostatic and magnetic fluctuations under conditions of combined inductive/LHCD regimes and of the relation of the level of these fluctuations to the anomalous particles transport in tokamak CASTOR. Results of the study are discussed in some detail. (author). 24 figs., 51 refs
International Nuclear Information System (INIS)
Kelly, J.L.; Rose, R.P.
1981-01-01
At a time when the potential benefits of various energy options are being seriously evaluated in many countries through-out the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode. (orig.)
Tokamak Plasmas : Mirnov coil data analysis for tokamak ADITYA
Indian Academy of Sciences (India)
The spatial and temporal structures of magnetic signal in the tokamak ADITYA is analysed using recently developed singular value decomposition (SVD) technique. The analysis technique is ﬁrst tested with simulated data and then applied to the ADITYA Mirnov coil data to determine the structure of current peturbation as ...
Radiation (UV-IR) measurements in the TRIAM-1 tokamak
Energy Technology Data Exchange (ETDEWEB)
Nakamura, K; Nakamura, Y; Hiraki, N; Itoh, S [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics
1981-07-01
Scaling law is presented that the ratio P sub(R)/P sub(I) (P sub(R): radiation loss, P sub(I): input power) is proportional to the parameter a sup(3/2)/R (a: minor radius, R: major radius). The discussion is presented about the design of the apparatus for measuring the radiation (ultraviolet to infrared) from the TRIAM-1 tokamak plasma.
Tokamak startup using point-source dc helicity injection.
Battaglia, D J; Bongard, M W; Fonck, R J; Redd, A J; Sontag, A C
2009-06-05
Startup of a 0.1 MA tokamak plasma is demonstrated on the ultralow aspect ratio Pegasus Toroidal Experiment using three localized, high-current density sources mounted near the outboard midplane. The injected open field current relaxes via helicity-conserving magnetic turbulence into a tokamaklike magnetic topology where the maximum sustained plasma current is determined by helicity balance and the requirements for magnetic relaxation.
Asymptotic stability boundaries of ballooning modes in circular tokamaks
International Nuclear Information System (INIS)
Chen, L.; Bondeson, A.; Chance, M.S.
1987-06-01
The model ballooning mode equation of Connor, Hastie, and Taylor for large-aspect-ratio circular tokamaks is analyzed in the limit of large pressure gradient, and corresponding expressions for stability boundaries are derived. In particular, it is found that for a fixed radial wave number, there exists an infinite sequence of unstable bands, and that minimizing over the radial wave numbers leads to asymptotic merging between the neighboring bands
Note for the Mirnov signal analysis in tokamaks
International Nuclear Information System (INIS)
Kikuchi, M.
1985-05-01
The relation between Mirnov coil signals and the current perturbation on the rational surface is examined analytically by using the approximate Green's function for the case of large aspect ratio circular tokamaks. Satellite island formation, phase modulation effect due to the poloidal variation of the field line pitch, and the shift effect of the plasma column with respect to the center of the vacuum chamber are examined. The detectability of these effects from Mirnov coil signals is discussed for TFTR
Instrumentation and controls of an ignited tokamak
International Nuclear Information System (INIS)
Becraft, W.R.; Golzy, J.; Houlberg, W.A.; Kukielka, C.A.; Onega, R.J.; Raju, G.V.S.; Stone, R.S.
1980-10-01
The instrumentation and controls (I and C) of an ignited plasma magnetically confined in a tokamak configuration needs increased emphasis in the following areas: (1) physics implications for control; (2) plasma shaping/position control; and (3) control to prevent disruptive instabilities. This document reports on the FY 1979 efforts in these and other areas. Also presented are discusssions in the areas of: (1) diagnostics suitable for the Engineering Test Facility (ETF); and (2) future research and development (R and D) needs. The appendices focus attention on some preliminary ideas about the measurement of the deuteron-triton (D-T) ratio in the plasma, synchrotron radiation, and divertor control. Finally, an appendix documenting the thermal consequences to the first wall of a MPD is presented
Coupling of tearing modes in tokamaks
International Nuclear Information System (INIS)
Finn, J.M.
1977-01-01
The simultaneous presence of tearing modes of different helical pitches leads to the destruction of magnetic surfaces, which has been suggested as the mechanism leading to the onset of the disruptive instability in tokamaks. For current profiles in which the m = 2 mode is unstable, but the m = 3 is stable, the coupling of the m = 3 to the m = 2 through the poloidal variation of the toroidal field can drive the m = 3 amplitude psi 3 to order psi 2 times the inverse aspect ratio. Detailed calculations, both analytical and numerical, have been performed for two models for the equilibrium and m = 2 mode structure. A slab model and incompressible m = 3 perturbations are assumed. The m = 3 amplitude increases with shear, up to a point, showing that as the current channel shrinks, overlap of resonances becomes more likely. The results also apply qualitatively to other m, m +- 1 interactions
Magnetic mapping of the TBR-1 tokamak
International Nuclear Information System (INIS)
Matta, Jose Antonio Sevidanes da.
1994-01-01
Axisymmetric hydromagnetic equilibria for an ideal conducting current fluid are described by means of an asymptotic expansion in powers of the inverse aspect ratio (ε α/R 0 ) that satisfies the Grad-Shafranov equation (equilibrium condition). The main profiles for these equilibria were computer. The effect of non-axisymmetric perturbations on the magnetic surfaces if the tokamak TBR-1 is investigated. The used method is able to show how magnetic field from external helical currents split the rational magnetic surface giving rise to the chain of magnetic islands. The Poincare map of the field line perturbed by resonant helical windings has been obtained numerically for the typical TBR-1 parameters. For increasing parameters of perturbation secondary resonances were observed, that transform the bound-state-like contours of a given island into similar structures of secondary magnetic islands. These results were used to determine the spectrum of the perturbation created by resonant helical windings. (author). 87 refs., 30 figs
Tearing mode analysis in tokamaks, revisited
International Nuclear Information System (INIS)
Nishimura, Y.; Callen, J.D.; Hegna, C.C.
1997-12-01
A new Δ' shooting code has been developed to investigate tokamak plasma tearing mode stability in a cylinder and large aspect ratio (ε ≤ 0.25) toroidal geometries, neglecting toroidal mode coupling. A different computational algorithm is used (shooting out from the singular surface instead of into it) to resolve the strong singularities at the mode rational surface, particularly in the presence of finite pressure term. Numerical results compare favorably with Furth et al. results. The effects of finite pressure, which are shown to decrease Δ', are discussed. It is shown that the distortion of the flux surfaces by the Shafranov shift, which modifies the geometry metric element stabilizes the tearing mode significantly, even in a low β regime before the toroidal magnetic curvature effects come into play. Double tearing modes in toroidal geometries are examined as well. Furthermore, m ≥ 2 tearing mode stability criteria are compared with three dimensional initial value MHD simulation by the FAR code
Compact tokamak reactors part 2 (numerical results)
International Nuclear Information System (INIS)
Wiley, J.C.; Wootton, A.J.; Ross, D.W.
1996-01-01
The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced
Energy losses on tokamak startup
International Nuclear Information System (INIS)
Murray, J.G.; Rothe, K.E.; Bronner, G.
1983-01-01
During the startup of a tokamak reactor using poloidal field (PF) coils to induce plasma currents, the conducting structures carry induced currents. The associated energy losses in the circuits must be provided by the startup coils and the PF system. This paper provides quantitative and comparitive values for the energies required as a function of the thickness or resistivity of the torus shells
Prospects for Tokamak Fusion Reactors
International Nuclear Information System (INIS)
Sheffield, J.; Galambos, J.
1995-01-01
This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant
Integral torque balance in tokamaks
International Nuclear Information System (INIS)
Pustovitov, V.D.
2011-01-01
The study is aimed at clarifying the balance between the sinks and sources in the problem of intrinsic plasma rotation in tokamaks reviewed recently by deGrassie (2009 Plasma Phys. Control. Fusion 51 124047). The integral torque on the toroidal plasma is calculated analytically using the most general magnetohydrodynamic (MHD) plasma model taking account of plasma anisotropy and viscosity. The contributions due to several mechanisms are separated and compared. It is shown that some of them, though, possibly, important in establishing the rotation velocity profile in the plasma, may give small input into the integral torque, but an important contribution can come from the magnetic field breaking the axial symmetry of the configuration. In tokamaks, this can be the error field, the toroidal field ripple or the magnetic perturbation created by the correction coils in the dedicated experiments. The estimates for the error-field-induced electromagnetic torque show that the amplitude of this torque is comparable to the typical values of torques introduced into the plasma by neutral beam injection. The obtained relations allow us to quantify the effect that can be produced by the existing correction coils in tokamaks on the plasma rotation, which can be used in experiments to study the origin and physics of intrinsic rotation in tokamaks. Several problems are proposed for theoretical studies and experimental tests.
ECRH Studies on Tokamak Plasmas.
1980-10-10
r.I*cru.Dtrtibution uUnliited 300 Unicorn Pork Drive Woburn, Massachusetts 04801 ECRH STUDIES ON TOKAMAK PLASMAS JAYCOR Project No. 6183 Final Report...up techniques now in use or being suggested, include growing the plasma from a small minor radius or applying a negative voltage spike immediately
Tokamak impurity-control techniques
International Nuclear Information System (INIS)
Schmidt, J.A.
1980-01-01
A brief review is given of the impurity-control functions in tokamaks, their relative merits and disadvantages and some prominent edge-interaction-control techniques, and there is a discussion of a new proposal, the particle scraper, and its potential advantages. (author)
An enhanced tokamak startup model
Goswami, Rajiv; Artaud, Jean-François
2017-01-01
The startup of tokamaks has been examined in the past in varying degree of detail. This phase typically involves the burnthrough of impurities and the subsequent rampup of plasma current. A zero-dimensional (0D) model is most widely used where the time evolution of volume averaged quantities determines the detailed balance between the input and loss of particle and power. But, being a 0D setup, these studies do not take into consideration the co-evolution of plasma size and shape, and instead assume an unchanging minor and major radius. However, it is known that the plasma position and its minor radius can change appreciably as the plasma evolves in time to fill in the entire available volume. In this paper, an enhanced model for the tokamak startup is introduced, which for the first time takes into account the evolution of plasma geometry during this brief but highly dynamic period by including realistic one-dimensional (1D) effects within the broad 0D framework. In addition the effect of runaway electrons (REs) has also been incorporated. The paper demonstrates that the inclusion of plasma cross section evolution in conjunction with REs plays an important role in the formation and development of tokamak startup. The model is benchmarked against experimental results from ADITYA tokamak.
Multimegawatt neutral beams for tokamaks
International Nuclear Information System (INIS)
Kunkel, W.B.
1979-03-01
Most of the large magnetic confinement experiments today and in the near future use high-power neutral-beam injectors to heat the plasma. This review briefly describes this remarkable technique and summarizes recent results as well as near term expectations. Progress has been so encouraging that it seems probable that tokamaks will achieve scientific breakeven before 1990
Joint research using small tokamaks
Czech Academy of Sciences Publication Activity Database
Gryaznevich, M.P.; Del Bosco, E.; Malaquias, A.; Mank, G.; Van Oost, G.; He, Yexi; Hegazy, H.; Hirose, A.; Hron, Martin; Kuteev, B.; Ludwig, G.O.; Nascimento, I.C.; Silva, C.; Vorobyev, G.M.
2005-01-01
Roč. 45, č. 10 (2005), S245-S254 ISSN 0029-5515. [Fusion Energy Conference contributions. Vilamoura, 1.11.2004-6.11.2004] Institutional research plan: CEZ:AV0Z20430508 Keywords : small tokamaks * thermonuclear fusion Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.418, year: 2005
Simulations of edge and scrape off layer turbulence in mega ampere spherical tokamak plasmas
DEFF Research Database (Denmark)
Militello, F; Fundamenski, W; Naulin, Volker
2012-01-01
The L-mode interchange turbulence in the edge and scrape-off-layer (SOL) of the tight aspect ratio tokamak MAST is investigated numerically. The dynamics of the boundary plasma are studied using the 2D drift-fluid code ESEL, which has previously shown good agreement with large aspect ratio machin...
On steady poloidal and toroidal flows in tokamak plasmas
International Nuclear Information System (INIS)
McClements, K. G.; Hole, M. J.
2010-01-01
The effects of poloidal and toroidal flows on tokamak plasma equilibria are examined in the magnetohydrodynamic limit. ''Transonic'' poloidal flows of the order of the sound speed multiplied by the ratio of poloidal magnetic field to total field B θ /B can cause the (normally elliptic) Grad-Shafranov (GS) equation to become hyperbolic in part of the solution domain. It is pointed out that the range of poloidal flows for which the GS equation is hyperbolic increases with plasma beta and B θ /B, thereby complicating the problem of determining spherical tokamak plasma equilibria with transonic poloidal flows. It is demonstrated that the calculation of the hyperbolicity criterion can be easily modified when the assumption of isentropic flux surfaces is replaced with the more tokamak-relevant one of isothermal flux surfaces. On the basis of the latter assumption, a simple expression is obtained for the variation of density on a flux surface when poloidal and toroidal flows are simultaneously present. Combined with Thomson scattering measurements of density and temperature, this expression could be used to infer information on poloidal and toroidal flows on the high field side of a tokamak plasma, where direct measurements of flows are not generally possible. It is demonstrated that there are four possible solutions of the Bernoulli relation for the plasma density when the flux surfaces are assumed to be isothermal, corresponding to four distinct poloidal flow regimes. Finally, observations and first principles-based theoretical modeling of poloidal flows in tokamak plasmas are briefly reviewed and it is concluded that there is no clear evidence for the occurrence of supersonic poloidal flows.
Current drive by Alfven waves in elongated cross section tokamak
International Nuclear Information System (INIS)
Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A.; Assis, A.S. de
1997-01-01
Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves 9GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory
Current drive by Alfven waves in elongated cross section tokamak
Energy Technology Data Exchange (ETDEWEB)
Tsypin, V.S.; Elfimov, A.G.; Nekrasov, F.M.; Azevedo, C.A. [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Fisica; Assis, A.S. de [Universidade Federal Fluminense, Niteroi, RJ (Brazil). Inst. de Fisica
1997-12-31
Full text. The problem of the noninductive current drive in cylindrical plasma model and in circular cross-section tokamaks had been already discussed intensively. At the beginning of the study of this problem it have been clear that there are significant difficulties in using of the current-drive in toroidal magnetic traps, especially in a tokamak reactor. Thus, in the case of the lower-hybrid current-drive the efficiency of this current-drive drops strongly as the plasma density increases. For the Alfven waves, there is an opinion that the efficiency of the current-drive drops as a result of waves absorption by the trapped particles 1,2. Okhawa proposed that the current in a magnetized plasma can be maintained also by means of forces, depending on the radiofrequency (rf) field amplitude gradients (the helicity injection). This idea was developed later, some new hopes appeared, connected with the possibility of the current-drive efficiency increasing. It was shown that for the cylindrical plasmas the local efficiency of Alfev wave current drive can be increased by one order of magnitude due to gradient forces, for the kinetic Alfven waves (KAW) and the global Alfven waves (GAW) at some range of the phase velocity. For tokamaks, this additional nonresonant current drive does not depend on the trapped particle effects, which reduce strongly the Alfven current drive efficiency in tokamaks, as it is supposed. Now, the theory development of the Alfven wave (AW) current drive is very important in the cource of the future experiments on the TCA/BR tokamak (Brazil). In this paper, an attempt is made to clarify some general aspects of this problems for magnetic traps. For large aspects ratio tokamaks, with an elongated cross-section, some general formulas concerning the untrapped and trapped particles dynamics and their input to the Landau damping of the Alfven waves, are presented. They are supposed to be used for the further development of the Alfven current drive theory
Advanced statistics for tokamak transport colinearity and tokamak to tokamak variation
International Nuclear Information System (INIS)
Riedel, K.S.
1989-01-01
This paper is an expository introduction to advanced statistics and scaling laws and their application to tokamak devices. Topics of discussion are as follows: implicit assumptions in the standard analysis; advanced regression techniques; specialized tools in statistics and their applications in fusion physics; and improved datasets for transport studies
Neoclassical dissipation and resistive wall modes in tokamaks
International Nuclear Information System (INIS)
Shaing, K.C.
2004-01-01
It is shown that the critical toroidal plasma flow speed that is required to stabilize the resistive wall mode in tokamaks is reduced by a factor of the order of B/B θ or of 1.265ε 3sol4 B/B θ depending on the plasma parameters when the perturbed neoclassical viscosity driven current is taken into account. Here, B is the magnetic field strength, B θ is the poloidal magnetic field strength, and ε is the inverse aspect ratio. This effect is illustrated using an existing model for the resistive wall modes by including the neoclassical dissipation in the derivation of the dispersion relation. The derivation is based on fluid equations with the plasma viscosity, calculated using kinetic equation, as the closure. The reduction of the critical toroidal speed is a consequence of the parallel (to the magnetic field B) momentum equation when neoclassical viscosity becomes important. The results are compared with experimental observations in tokamaks
Plasma shaping effects on tokamak scrape-off layer turbulence
Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo
2017-03-01
The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\
Multi-channel bolometer system on JFT-2M tokamak
International Nuclear Information System (INIS)
Tamai, Hiroshi; Maeno, Masaki; Matsuda, Toshiaki; Matoba, Tohru
1988-07-01
Multi-channel bolometer system is designed and installed to observe the radiation profile on JFT-2M tokamak. Sensor head is made of Thinistor, which is a kind of semiconductor, because it has the advantage of higher sensitivity of about one order of magnitude than the conventional metal foil bolometer and is suitable for the profile measurement in which the signal from the plasma is relatively small. The response and cooling characteristics of the bolometer sensor are suitable for the condition of JFT-2M tokamak plasma. Low noise circuit of bridge and differentiator is developed to optimize the signal to noise ratio in the JFT-2M operating condition. With use of the bolometer system, the radiation profile in joule heating plasma as well as additional heating plasma especially in H-mode plasma is successfully observed. (author)
Conceptual design of a Tokamak hybrid power reactor (THPR)
International Nuclear Information System (INIS)
Matsuoka, F.; Imamura, Y.; Inoue, M.; Asami, N.; Kasai, M.; Yanagisawa, I.; Ida, T.; Takuma, T.; Yamaji, K.; Akita, S.
1987-01-01
A conceptual design of a fusion-fission hybrid tokamak reactor has been carried out to investigate the engineering feasibility and promising scale of a commercial hybrid reactor power plant. A tokamak fusion driver based on the recent plasma scaling law is introduced in this design study. The major parameters and features of the reactor are R=6.06 m, a=1.66 m, Ip=11.8 MA, Pf=668 MW, double null divertor plasma and steady state burning with RF current drive. The fusion power has been determined with medium energy multiplication in the blanket so as to relieve thermal design problems and produce electric power around 1000 MW. Uranium silicide is used for the fast fission blanket material to promise good nuclear performance. The coolant of the blanket is FLIBE and the tritium breeding blanket material is Li 2 O ceramics providing breeding ratio above unity
Microwave Tokamak Experiment: Overview and status
International Nuclear Information System (INIS)
1990-05-01
The Microwave Tokamak Experiment, now under construction at the Laboratory, will use microwave heating from a free-electron laser. The intense microwave pulses will be injected into the tokamak to realize several goals, including a demonstration of the effects of localized heat deposition within magnetically confined plasma, a better understanding of energy confinement in tokamaks, and use of the new free-electron laser technology for plasma heating. 3 figs., 3 tabs
Combined confinement system applied to tokamaks
International Nuclear Information System (INIS)
Ohkawa, Tihiro
1986-01-01
From particle orbit point of view, a tokamak is a combined confinement configuration where a closed toroidal volume is surrounded by an open confinement system like a magnetic mirror. By eliminating a cold halo plasma, the energy loss from the plasma becomes convective. The H-mode in diverted tokamaks is an example. Because of the favorable scaling of the energy confinement time with temperature, the performance of the tokamak may be significantly improved by taking advantage of this effect. (author)
Presheath profiles in simulated tokamak edge plasmas
International Nuclear Information System (INIS)
LaBombard, B.; Conn, R.W.; Hirooka, Y.; Lehmer, R.; Leung, W.K.; Nygren, R.E.; Ra, Y.; Tynan, G.
1988-04-01
The PISCES plasma surface interaction facility at UCLA generates plasmas with characteristics similar to those found in the edge plasmas of tokamaks. Steady state magnetized plasmas produced by this device are used to study plasma-wall interaction phenomena which are relevant to tokamak devices. We report here progress on some detailed investigations of the presheath region that extends from a wall surface into these /open quotes/simulated tokamak/close quotes/ edge plasma discharges along magnetic field lines
Noninductively Driven Tokamak Plasmas at Near-Unity Toroidal Beta
International Nuclear Information System (INIS)
Schlossberg, David J.; Bodner, Grant M.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.
2017-01-01
Access to and characterization of sustained, toroidally confined plasmas with a very high plasma-to-magnetic pressure ratio (β t ), low internal inductance, high elongation, and nonsolenoidal current drive is a central goal of present tokamak plasma research. Stable access to this desirable parameter space is demonstrated in plasmas with ultralow aspect ratio and high elongation. Local helicity injection provides nonsolenoidal sustainment, low internal inductance, and ion heating. Equilibrium analyses indicate β t up to ~100% with a minimum |B| well spanning up to ~50% of the plasma volume.
Modelling and control of a tokamak plasma; Modelisation et commande d`un plasma de tokamak
Energy Technology Data Exchange (ETDEWEB)
Bremond, S
1995-10-18
Vertically elongated tokamak plasmas, while attractive as regards Lawson criteria, are intrinsically instable. It is found that the open-loop instability dynamics is characterised by the relative value of two dimensionless parameters: the coefficient of inductive coupling between the vessel and the coils, and the coil damping efficiency on the plasma displacement relative to that of the vessel. Applications to Tore Supra -where the instability is due to the iron core attraction- and DIII-D are given. A counter-effect of the vessel, which temporarily reverses the effect of coil control on the plasma displacement, is seen when the inductive coupling is higher than the damping ratio. Precise control of the plasma boundary is necessary if plasma-wall interaction and/or coupling to heating antennas are to be monitored. A positional drift, of a few mm/s, which had been observed in the Tore Supra tokamak, is explained and corrected. A linear plasma shape response model is then derived from magnetohydrodynamic equilibrium calculation, and proved to be in good agreement with experimental data. An optimal control law is derived, which minimizes an integral quadratic criteria on tracking errors and energy expenditure. This scheme avoids compensating coil currents, and could render local plasma shaping more precise. (authors). 123 refs., 77 figs., 6 tabs., 4 annexes.
Improvement of the tokamak concept
Energy Technology Data Exchange (ETDEWEB)
Laurent, L
1994-12-31
Improvement of the tokamak concept is highly desirable to reduce the size and capital cost of a device able to ignite to increase the plasma pressure, i.e. the power density to reduce the cost of electricity, and to increase the fraction of bootstrap current to render the tokamak compatible with continuous operation. The most important results obtained in this field are summarized, and the options are shown which are still open and explored by the various experiments. Various effects of the plasma shaping are discussed, plasma configurations with both high {beta}{sub N} and H{sub G} are explored, and the issues of stable steady state and of the plasma edge are briefly discussed. (R.P.). 65 refs., 2 tabs.
Flux driven turbulence in tokamaks
International Nuclear Information System (INIS)
Garbet, X.; Ghendrih, P.; Ottaviani, M.; Sarazin, Y.; Beyer, P.; Benkadda, S.; Waltz, R.E.
1999-01-01
This work deals with tokamak plasma turbulence in the case where fluxes are fixed and profiles are allowed to fluctuate. These systems are intermittent. In particular, radially propagating fronts, are usually observed over a broad range of time and spatial scales. The existence of these fronts provide a way to understand the fast transport events sometimes observed in tokamaks. It is also shown that the confinement scaling law can still be of the gyroBohm type in spite of these large scale transport events. Some departure from the gyroBohm prediction is observed at low flux, i.e. when the gradients are close to the instability threshold. Finally, it is found that the diffusivity is not the same for a turbulence calculated at fixed flux than at fixed temperature gradient, with the same time averaged profile. (author)
Options for an ignited tokamak
International Nuclear Information System (INIS)
Sheffield, J.
1984-02-01
It is expected that the next phase of the fusion program will involve a tokamak with the goals of providing an ignited plasma for pulses of hundreds of seconds. A simple model is described in this memorandum which establishes the physics conditions for such a self-sustaining plasma, for given ion and electron thermal diffusivities, in terms of R/a, b/a, I, B/q, epsilon β/sub p/, anti T/sub i/, and anti T/sub e//anti T/sub i/. The model is used to produce plots showing the wide range of tokamaks that may ignite or have a given ignition margin. The constraints that limit this range are discussed
Plasma diagnostics on large tokamaks
International Nuclear Information System (INIS)
Orlinskij, D.V.; Magyar, G.
1988-01-01
The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs
Starfire: a commercial tokamak reactor
International Nuclear Information System (INIS)
Baker, C.C.; Abdou, M.A.; DeFreece, D.A.; Trachsel, C.A.; Graumann, D.; Kokoszenski, J.
1979-01-01
The basic objective of the STARFIRE Project is to develop a design concept for a commercial tokamak fusion electric power plant based on the deuterium/tritium/lithium fuel cycle. The key technical objective is to develop the best embodiment of the tokamak as a power reactor consistent with credible engineering solutions to design problems. Another key goal of the project is to give careful attention to the safety and environmental features of a commercial fusion reactor. The STARFIRE Project was initiated in May 1979, with the goal of completing the design study by October 1980. The purpose of this paper is to present an overview of the major parameters and design features that have been tentatively selected for STARFIRE
Comprehensive numerical modelling of tokamaks
International Nuclear Information System (INIS)
Cohen, R.H.; Cohen, B.I.; Dubois, P.F.
1991-01-01
We outline a plan for the development of a comprehensive numerical model of tokamaks. The model would consist of a suite of independent, communicating packages describing the various aspects of tokamak performance (core and edge transport coefficients and profiles, heating, fueling, magnetic configuration, etc.) as well as extensive diagnostics. These codes, which may run on different computers, would be flexibly linked by a user-friendly shell which would allow run-time specification of packages and generation of pre- and post-processing functions, including workstation-based visualization of output. One package in particular, the calculation of core transport coefficients via gyrokinetic particle simulation, will become practical on the scale required for comprehensive modelling only with the advent of teraFLOP computers. Incremental effort at LLNL would be focused on gyrokinetic simulation and development of the shell
Magnetic island formation in tokamaks
International Nuclear Information System (INIS)
Yoshikawa, S.
1989-04-01
The size of a magnetic island created by a perturbing helical field in a tokamak is estimated. A helical equilibrium of a current- carrying plasma is found in a helical coordinate and the helically flowing current in the cylinder that borders the plasma is calculated. From that solution, it is concluded that the helical perturbation of /approximately/10/sup /minus/4/ of the total plasma current is sufficient to cause an island width of approximately 5% of the plasma radius. 6 refs
Equilibrium Reconstruction in EAST Tokamak
International Nuclear Information System (INIS)
Qian Jinping; Wan Baonian; Shen Biao; Sun Youwen; Liu Dongmei; Xiao Bingjia; Ren Qilong; Gong Xianzu; Li Jiangang; Lao, L. L.; Sabbagh, S. A.
2009-01-01
Reconstruction of experimental axisymmetric equilibria is an important part of tokamak data analysis. Fourier expansion is applied to reconstruct the vessel current distribution in EFIT code. Benchmarking and testing calculations are performed to evaluate and validate this algorithm. Two cases for circular and non-circular plasma discharges are presented. Fourier expansion used to fit the eddy current is a robust method and the real time EFIT can be introduced to the plasma control system in the coming campaign. (magnetically confined plasma)
Relaxed states of tokamak plasmas
International Nuclear Information System (INIS)
Kucinski, M.Y.; Okano, V.
1993-01-01
The relaxed states of tokamak plasmas are studied. It is assumed that the plasma relaxes to a quasi-steady state which is characterized by a minimum entropy production rate, compatible with a number of prescribed conditions and pressure balance. A poloidal current arises naturally due to the anisotropic resistivity. The minimum entropy production theory is applied, assuming the pressure equilibrium as fundamental constraint on the final state. (L.C.J.A.)
Runaway electrons during tokamak startup
International Nuclear Information System (INIS)
Sharma, A.S.; Jayakumar, R.
1988-01-01
Runaway electrons significantly affect the plasma and impurity evolution during tokamak startup. During its rise, a runaway pulse stores magnetic flux inductively; this is then released during the decay phase of the runaway pulse. This process affects plasma formation, current initiation and current buildup. Because of their relativistic velocities the runaway electrons have higher ionization and excitation rates than the plasma electrons. This leads to a significant modification of the impurity behaviour and consequently the plasma evolution. (author). 20 refs, 8 figs
Minimum scaling laws in tokamaks
International Nuclear Information System (INIS)
Zhang, Y.Z.; Mahajan, S.M.
1986-10-01
Scaling laws governing anomalous electron transport in tokamaks with ohmic and/or auxiliary heating are derived using renormalized Vlasov-Ampere equations for low frequency electromagnetic microturbulence. It is also shown that for pure auxiliary heating (or when auxiliary heating power far exceeds the ohmic power), the energy confinement time scales as tau/sub E/ ∼ P/sub inj//sup -1/3/, where P/sub inj/ is the injected power
Gyrosheath near the tokamak edge
International Nuclear Information System (INIS)
Hazeltine, R.D.; Xiao, H.; Valanju, P.M.
1993-03-01
A new model for the structure of the radial electric field profile in the edge during the H-mode is proposed. Charge separation caused by the difference between electron and ion gyromotion, or more importantly in a tokamak, the banana motion (halo effect) can self-consistently produce an electric dipole moment that causes the sheared radial electric field. The calculated results based on the model are consistent with D-III D and TEXTOR experimental results
Tokamak plasma boundary layer model
International Nuclear Information System (INIS)
Volkov, T.F.; Kirillov, V.D.
1983-01-01
A model has been developed for the limiter layer and for the boundary region of the plasma column in a tokamak to facilitate analytic calculations of the thickness of the limiter layers, the profiles and boundary values of the temperature and the density under various conditions, and the difference between the electron and ion temperatures. This model can also be used to analyze the recycling of neutrals, the energy and particle losses to the wall and the limiter, and other characteristics
Shear Alfven waves in tokamaks
International Nuclear Information System (INIS)
Kieras, C.E.
1982-12-01
Shear Alfven waves in an axisymmetric tokamak are examined within the framework of the linearized ideal MHD equations. Properties of the shear Alfven continuous spectrum are studied both analytically and numerically. Implications of these results in regards to low frequency rf heating of toroidally confined plasmas are discussed. The structure of the spatial singularities associated with these waves is determined. A reduced set of ideal MHD equations is derived to describe these waves in a very low beta plasma
Discharge cleaning for a tokamak
International Nuclear Information System (INIS)
Ishii, Shigeyuki
1983-01-01
Various methods of discharge cleaning for tokamaks are described. The material of the first walls of tokamaks is usually stainless steel, inconel, titanium and so on. Hydrogen is exclusively used as the discharge gas. Glow discharge cleaning (GDC), Taylor discharge cleaning (TDC), and electron cyclotron resonance discharge cleaning (ECR-DC) are discussed in this paper. The cleaning by GDC is made by moving a movable anode to the center of a tokamak vassel. Taylor found the good cleaning effect of induced discharge by high pressure and low power discharge. This is called TDC. When the frequency of high frequency discharge in a magnetic field is equal to that of the electron cyclotron resonance, the break down potential is lowered if the pressure is sufficiently low. The ECR-CD is made by using this effect. In TDC and ECR-DC, the electron temperature, which has a close relation to the production rate of H 0 , can be controlled by the pressure. In GDC, the operating pressure was improved by the radio frequency glow (RG) method. However, there is still the danger of arcing. In case of GDC and ECR-DC, the position of plasma can be controlled, but not in case of TDC. The TDC is accepted by most of takamak devices in the world. (Kato, T.)
Magnetic confinement experiment. I: Tokamaks
International Nuclear Information System (INIS)
Goldston, R.J.
1995-08-01
Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices
Transcription and the aspect ratio of DNA
DEFF Research Database (Denmark)
Olsen, Kasper Wibeck; Bohr, Jakob
2013-01-01
analysis of transcription. It is shown that under certain reasonable assumptions transcription is only possible if the aspect ratio is in the regime corresponding to further twisting. We find this constraint to be in agreement with long-established crystallographic studies of DNA.......Two separate regimes exist for the aspect ratio of DNA. A low aspect regime where DNA will twist further under strain and a high aspect regime where DNA will untwist under strain. The question of the overall geometry, i.e. the aspect ratio, of DNA is revisited from the perspective of a geometrical...
Zonal flows in tokamaks with anisotropic pressure
International Nuclear Information System (INIS)
Ren, Haijun
2014-01-01
Zonal flows (ZFs) in a tokamak plasma with anisotropic pressure are investigated. The dynamics of perpendicular and parallel pressures are determined by the Chew-Goldberger-Low double equations and low-β condition is adopted, where β is the ratio of plasma pressure to the magnetic field pressure. The dispersion relation is analytically derived and illustrates two branches of ZFs. The low frequency zonal flow (LFZF) branch becomes unstable when χ, the ratio of the perpendicular pressure to the parallel one, is greater than a threshold value χ c , which is about 3.8. In the stable region, its frequency increases first and then decreases with increasing χ. For χ = 1, the frequency of LFZF agrees well with the experimental observation. For the instability, the growth rate of LFZF increases with χ. The geodesic acoustic mode branch is shown to be always stable with a frequency increasing with χ. The safety factor is shown to diminish the frequencies of both branches or the growth rate of LFZF
Non-Axisymmetric Shaping of Tokamaks Preserving Quasi-Axisymmetry
Energy Technology Data Exchange (ETDEWEB)
Long-Poe Ku and Allen H. Boozer
2009-06-05
If quasi-axisymmetry is preserved, non-axisymmetric shaping can be used to design tokamaks that do not require current drive, are resilient to disruptions, and have robust plasma stability without feedback. Suggestions for addressing the critical issues of tokamaks can only be validated when presented with sufficient specificity that validating experiments can be designed. The purpose of this paper is provide that specificity for non-axisymmetric shaping. To our knowledge, no other suggestions for the solution of a number of tokamak issues, such as disruptions, have reached this level of specificity. Sequences of three-field-period quasi-axisymmetric plasmas are studied. These sequences address the questions: (1) What can be achieved at various levels of non-axisymmetric shaping? (2) What simplifications to the coils can be achieved by going to a larger aspect ratio? (3) What range of shaping can be achieved in a single experimental facility? The sequences of plasmas found in this study provide a set of interesting and potentially important configurations.
Magnetic confinement by Tokamak: physical aspects
International Nuclear Information System (INIS)
Tachon, J.
1980-01-01
After describing the Tokamak configuration concept, the author provides an analysis of the principal physical aspects of this type of installation and concludes by estimating that the Tokamak concept is a 'plausible candidate' as a means of producing controlled thermonuclear fusion [fr
Economic evaluation of tokamak power plants
International Nuclear Information System (INIS)
Reid, R.L.; Steiner, D.
1977-01-01
This study reports the impact of plasma operating characteristics, engineering options, and technology on the capital cost trends of tokamak power plants. Tokamak power systems are compared to other advanced energy systems and found to be economically competitive. A three-phase strategy for demonstrating commercial feasibility of fusion power, based on a common-site multiple-unit concept, is presented
Simulation of a major tokamak disruption
International Nuclear Information System (INIS)
White, R.B.; Monticello, D.A.; Rosenbluth, M.N.
1977-08-01
It is known that the internal tokamak disruption leads to a current profile which is flattened inside the surface where the safety factor equals unity. It is shown that such a profile can lead to m = 2 magnetic islands which grow to fill a substantial part of the tokamak cross section in a time consistent with the observations of the major disruption
Diagnostics for the Rijnhuizen Tokamak Project
Donne, A. J. H.
1994-01-01
The research program of the Rijnhuizen Tokamak Project is concentrated on the study of plasma transport processes. The RTP tokamak is therefore equipped with an extensive set of multichannel diagnostics, including a 19-channel FIR interferometer, a 20-channel heterodyne ECE system, an 80-channel
Mercier criterion for high-β tokamaks
International Nuclear Information System (INIS)
Galvao, R.M.O.
1984-01-01
An expression, for the application of the Mercier criterion to numerical studies of diffuse high-β tokamaks (β approximatelly Σ,q approximatelly 1), which contains only leading order contributions in the high-β tokamak approximation is derived. (L.C.) [pt
The ARIES-I tokamak reactor study
International Nuclear Information System (INIS)
1991-01-01
This report contains an overview of the Aries-I tokamak reactor study. The following topics are discussed on this tokamak: Systems studies; equilibrium, stability, and transport; summary and conclusions; current drive; impurity control system; tritium systems; magnet engineering; fusion-power-core engineering; power conversion; Aries-I safety design and analysis; design layout and maintenance; and start-up and operations
Engineering Design of KSTAR tokamak main structure
International Nuclear Information System (INIS)
Im, K.H.; Cho, S.; Her, N.I.
2001-01-01
The main components of the KSTAR (Korea Superconducting Tokamak Advanced Research) tokamak including vacuum vessel, plasma facing components, cryostat, thermal shield and magnet supporting structure are in the final stage of engineering design. Hundai Heavy Industries (HHI) has been involved in the engineering design of these components. The current configuration and the final engineering design results for the KSTAR main structure are presented. (author)
Summary report on tokamak confinement experiments
International Nuclear Information System (INIS)
1982-03-01
There are currently five major US tokamaks being operated and one being constructed under the auspices of the Division of Toroidal Confinement Systems. The currently operating tokamaks include: Alcator C at the Massachusetts Institute of Technology, Doublet III at the General Atomic Company, the Impurity Studies Experiment (ISX-B) at the Oak Ridge National Laboratory, and the Princeton Large Torus (PLT) and the Poloidal Divertor Experiment (PDX) at the Princeton Plasma Physics Laboratory. The Tokamak Fusion Test Reactor (TFTR) is under construction at Princeton and should be completed by December 1982. There is one major tokamak being funded by the Division of Applied Plasma Physics. The Texas Experimental Tokamak (TEXT) is being operated as a user facility by the University of Texas. The TEXT facility includes a complete set of standard diagnostics and a data acquisition system available to all users
Energy Technology Data Exchange (ETDEWEB)
Valencia A, R
2006-07-01
spectroscopy and plasma resistivity measurement. Such mechanism depends fundamentally on the mass of the ions that interact with the wall during the plasma current formation phase. The reaction products generated by the glow discharge plasma, created in a calibrated mixture of methane-hydrogen during the hydrogenated amorphous carbon film deposit on the vessel wall of Novillo tokamak, were determined by mass spectrometry. By way of measuring the emission lines of the carbon and oxygen impurities in intense discharges, the time required by the plasma to interact with the wall was estimated. In addition to it, the temporal conduct of the emission line intensity of these impurities was observed by means of an intensified CCD detector. Once an {approx} 10 % of helium was introduced in the operating gas of the tokamak discharges, a 25-42 eV time variation of the electron temperature was measured using the intensity ratio technique. (Author)
Indian Academy of Sciences (India)
Our attraction to another body increases if the body is symmetricaland in proportion. If a face or a structure is in proportion,we are more likely to notice it and find it beautiful.The universal ratio of beauty is the 'Golden Ratio', found inmany structures. This ratio comes from Fibonacci numbers.In this article, we explore this ...
Indian Academy of Sciences (India)
Keywords. Fibonacci numbers, golden ratio, Sanskrit prosody, solar panel. Abstract. Our attraction to another body increases if the body is symmetricaland in proportion. If a face or a structure is in proportion,we are more likely to notice it and find it beautiful.The universal ratio of beauty is the 'Golden Ratio', found inmany ...
Indian Academy of Sciences (India)
Our attraction to another body increases if the body is sym- metrical and in proportion. If a face or a structure is in pro- portion, we are more likely to notice it and find it beautiful. The universal ratio of beauty is the 'Golden Ratio', found in many structures. This ratio comes from Fibonacci numbers. In this article, we explore this ...
Cluster storage for COMPASS tokamak
Czech Academy of Sciences Publication Activity Database
Písačka, Jan; Hron, Martin; Janky, Filip; Pánek, Radomír
2012-01-01
Roč. 87, č. 12 (2012), s. 2238-2241 ISSN 0920-3796. [IAEA Technical Meeting on Control, Data Acquisition, and Remote Participation for Fusion Research/8./. San Francisco, 20.06.2011-24.06.2011] R&D Projects: GA ČR GAP205/11/2470; GA MŠk 7G10072; GA MŠk(CZ) LM2011021 Institutional research plan: CEZ:AV0Z20430508 Keywords : COMPASS * Tokamak * Codac * Cluster * GlusterFS * Storage Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 0.842, year: 2012 http://dx.doi.org/10.1016/j.fusengdes.2012.09.006
Surface tearing modes in tokamaks
International Nuclear Information System (INIS)
Takizuka, Tomonori; Kurita, Gen-ichi; Azumi, Masafumi; Takeda, Tatsuoki
1985-10-01
Surface tearing modes in tokamaks are studied numerically and analytically. The eigenvalue problem is solved to obtain the growth rate and the mode structure. We investigate in detail dependences of the growth rate of the m/n = 2/1 resistive MHD modes on the safety factor at the plasma surface, current profile, wall position, and resistivity. The surface tearing mode moves the plasma surface even when the wall is close to the surface. The stability diagram for these modes is presented. (author)
Major disruption process in tokamak
International Nuclear Information System (INIS)
Kurita, Gen-ichi; Azumi, Masafumi; Tuda, Takashi; Takizuka, Tomonori; Tsunematsu, Toshihide; Tokuda, Shinji; Itoh, Kimitaka; Takeda, Tatsuoki
1981-11-01
The major disruption in a cylindrical tokamak is investigated by using the multi-helicity code, and the destabilization of the 3/2 mode by the mode coupling with the 2/1 mode is confirmed. The evolution of the magnetic field topology caused by the major disruption is studied in detail. The effect of the internal disruption on the 2/1 magnetic island width is also studied. The 2/1 magnetic island is not enhanced by the flattening of the q-profile due to the internal disruption. (author)
Modular coils and finite-β operation of a quasi-axially symmetric tokamak
International Nuclear Information System (INIS)
Drevlak, M.
1998-01-01
Quasi-axially symmetric tokamaks (QA tokamaks) are an extension of the conventional tokamak concept. In these devices the magnetic field strength is independent of the generalized toroidal magnetic co-ordinate even though the cross-sectional shape changes. An optimized plasma equilibrium belonging to the class of QA tokamaks has been proposed by Nuehrenberg. It features the small aspect ratio of a tokamak while allowing part of the rotational transform to be generated by the external field. In this article, two particular aspects of the viability of QA tokamaks are explored, namely the feasibility of modular coils and the possibility of maintaining quasi-axial symmetry in the free-boundary equilibria obtained with the coils found. A set of easily feasible modular coils for the configuration is presented. It was designed using the extended version of the NESCOIL code (MERKEL, P., Nucl. Fusion 27 (1987) 867). Using this coil system, free-boundary calculations of the plasma equilibrium were carried out using the NEMEC code (HIRSHMAN, S.P., VAN RIJ, W.I., MERKEL, P., Comput. Phys. Commun. 43 (1986) 143). It is observed that the effects of finite β and net toroidal plasma current can be compensated for with good precision by applying a vertical magnetic field and by separately adjusting the currents of the modular coils. A set of fully three dimensional (3-D) auxiliary coils is proposed to exert control on the rotational transform in the plasma. Deterioration of the quasi-axial symmetry induced by the auxiliary coils can be avoided by adequate adjustment of the currents in the primary coils. Finally, the neoclassical transport properties of the configuration are examined. It is observed that optimization with respect to confinement of the alpha particles can be maintained at operation with finite toroidal current if the aforementioned corrective measures are used. In this case, the neoclassical behaviour is shown to be very similar to that of a conventional tokamak
Impact of maximum TF magnetic field on performance and cost of an advanced physics tokamak
International Nuclear Information System (INIS)
Reid, R.L.
1983-01-01
Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of variation in the maximum value of the field at the toroidal field (TF) coils on the performance and cost of a low q/sub psi/, quasi-steady-state tokamak. Marginal ignition, inductive current startup plus 100 s of inductive burn, and a constant value of epsilon (inverse aspect ratio) times beta poloidal were global conditions imposed on this study. A maximum TF field of approximately 10 T was found to be appropriate for this device
Three-dimensional tokamak equilibria and stellarators with two-dimensional magnetic symmetry
International Nuclear Information System (INIS)
Garabedian, P.R.
1997-01-01
Three-dimensional computer codes have been developed to simulate equilibrium, stability and transport in tokamaks and stellarators. Bifurcated solutions of the tokamak problem suggest that three-dimensional effects may be more important than has generally been thought. Extensive calculations have led to the discovery of a stellarator configuration with just two field periods and with aspect ratio 3.2 that has a magnetic field spectrum B mn with toroidal symmetry. Numerical studies of equilibrium, stability and transport for this new device, called the Modular Helias-like Heliac 2 (MHH2), will be presented. (author)
Impact of major design parameters on the economics of Tokamak power plants
International Nuclear Information System (INIS)
Abdou, M.A.; Ehst, D.; Maroni, V.; Stacey, W.M. Jr.
1977-11-01
A parametric systems studies program is now in an active stage at Argonne National Laboratory. This paper presents a summary of results from this systems analysis effort. The impact of major design parameters on the economics of tokamak power plants is examined. The major parameters considered are: (1) the plant power rating; (2) toroidal-field strength; (3) plasma β/sub t/; (4) aspect ratio; (5) plasma elongation; (6) inner blanket/shield thickness; and (7) neutron wall load. The performance characteristics and economics of tokamak power plants are also compared for two structural materials
Energy confinement of tokamak plasma with consideration of bootstrap current effect
International Nuclear Information System (INIS)
Yuan Ying; Gao Qingdi
1992-01-01
Based on the η i -mode induced anomalous transport model of Lee et al., the energy confinement of tokamak plasmas with auxiliary heating is investigated with consideration of bootstrap current effect. The results indicate that energy confinement time increases with plasma current and tokamak major radius, and decreases with heating power, toroidal field and minor radius. This is in reasonable agreement with the Kaye-Goldston empirical scaling law. Bootstrap current always leads to an improvement of energy confinement and the contraction of inversion radius. When γ, the ratio between bootstrap current and total plasma current, is small, the part of energy confinement time contributed from bootstrap current will be about γ/2
Initial plasma production by induction electric field on QUEST tokamak
International Nuclear Information System (INIS)
Hasegawa, Makoto; Nakamura, Kazuo; Sato, Kohnosuke
2007-01-01
Induction electric field by center solenoid coil plays a roll to produce initial plasma. According to Townsend avalanche theory, minimum electric field for plasma breakdown depends on neutral gas pressure and connection length. On QUEST spherical tokamak, a connection length is evaluated as 966m on null point neighborhood with coil current ratio I PF26 /I CS =0.1, and induction electric field considering eddy current of vacuum vessel is evaluated as about 0.1 V/m on null point neighborhood. With Townsend avalanche theory, these values manage to produce initial plasma on QUEST. (author)
Resonant fields created by spiral electric currents in Tokamaks
International Nuclear Information System (INIS)
Fernandes, A.S.; Caldas, I.L.
1985-01-01
The influence of the resonant magnetic perturbations, created by electric currents in spirals, on the plasma confinement in a tokamak with circular section and large aspect ratio is investigated. These perturbations create magnetic islands around the rational magnetic surface which has the helicity of the helicoidal currents. The intensities of these currents are calculated in order to the magnetic islands reach the limiter or others rational surfaces, what could provoke the plasma disrupture. The electric current intensities are estimated, in two spiral sets with different helicities, which create a predominantly stocastic region among the rational magnetic surfaces with these helicities. (L.C.) [pt
Nuclear fusion research at Tokamak Energy Ltd
International Nuclear Information System (INIS)
Windridge, Melanie J.; Gryaznevich, Mikhail; Kingham, David
2017-01-01
Tokamak Energy's approach is close to the mainstream of nuclear fusion, and chooses a spherical tokamak, which is an economically developed form of Tokamak reactor design, as research subjects together with a high-temperature superconducting magnet. In the theoretical prediction, it is said that spherical tokamak can make tokamak reactor's scale compact compared with ITER or DEMO. The dependence of fusion energy multiplication factor on reactor size is small. According to model studies, it has been found that the center coil can be protected from heat and radiation damage even if the neutron shielding is optimized to 35 cm instead of 1 m. As a small tokamak with a high-temperature superconducting magnet, ST25 HTS, it demonstrated in 2015 continuous operation for more than 24 hours as a world record. Currently, this company is constructing a slightly larger ST40 type, and it is scheduled to start operation in 2017. ST40 is designed to demonstrate that it can realize a high magnetic field with a compact size and aims at attaining 8-10 keV (reaching the nuclear fusion reaction temperature at about 100 million degrees). This company will verify the startup and heating technology by the coalescence of spherical tokamak expected to have plasma current of 2 MA, and will also use 2 MW of neutral particle beam heating. In parallel with ST40, it is promoting a development program for high-temperature superconducting magnet. (A.O.)
Enhancement of confinement in tokamaks
International Nuclear Information System (INIS)
Furth, H.P.
1986-05-01
A plausible interpretation of the experimental evidence is that energy confinement in tokamaks is governed by two separate considerations: (1) the need for resistive MHD kink-stability, which limits the permissible range of current profiles - and therefore normally also the range of temperature profiles; and (2) the presence of strongly anomalous microscopic energy transport near the plasma edge, which calibrates the amplitude of the global temperature profile, thus determining the energy confinement time tau/sub E/. Correspondingly, there are two main paths towards the enhancement of tokamak confinement: (1) Configurational optimization, to increase the MHD-stable energy content of the plasma core, can evidently be pursued by varying the cross-sectional shape of the plasma and/or finding stable radial profiles with central q-values substantially below unity - but crossing from ''first'' to ''second'' stability within the peak-pressure region would have the greatest ultimate potential. (2) Suppression of edge turbulence, so as to improve the heat insulation in the outer plasma shell, can be pursued by various local stabilizing techniques, such as use of a poloidal divertor. The present confinement model and initial TFTR pellet-injection results suggest that the introduction of a super-high-density region within the plasma core should be particularly valuable for enhancing ntau/subE/. In D-T operation, a centrally peaked plasma pressure profile could possibly lend itself to alpha-particle-driven entry into the second-stability regime
International Nuclear Information System (INIS)
Greenspan, E.; Blue, T.; Miley, G.H.
1981-01-01
The domains of plasma fuel cycles bounded by the D-T and Cat-D, and by the D-T and SCD modes of operation are examined. These domains, referred to as, respectively, the Cat-D-T and SCD-T modes of operation, are characterized by the number (γ) of tritons per fusion neutron available from external (to the plasma) sources. Two external tritium sources are considered - the blankets of the Cat-D-T (SCD-T) reactors and fission reactors supported by the Cat-D-T (SCD-T) driven hybrid reactors. It is found that by using 6 Li for the active material of the control elements of the fission reactors, it is possible to achieve γ values close to unity. Cat-D-T tokamaks could be designed to have smaller size, higher power density, lower magnetic field and even lower plasma temperature than Cat-D tokamaks; the difference becomes significant for γ greater than or equal to .75. The SCD-T mode of operation appears to be even more attractive. Promising applications identified for these Cat-D-T and SCD-T modes of operation include hybrid reactors, fusion synfuel factories and fusion reactors which have difficulty in providing all their tritium needs
Power deposition for ion cyclotron heating in large tokamaks
International Nuclear Information System (INIS)
Hellsten, T.; Villard, L.
1988-01-01
The power deposition profiles during minority ion cyclotron heating are analysed in large tokamaks by using the global, toroidal wave code LION. For tokamaks with large aspect ratio and with circular cross-section, the wave is focused on the magnetic axis and can be absorbed there by cyclotron absorption when the cyclotron resonance passes through the magnetic axis. The power deposition profile is then essentially determined by the Doppler broadening of the ion cyclotron resonance. For equilibria either non-circular or with a small aspect ratio the power deposition profile depends also on the strength of the damping. In this case the power deposition profile can be expressed as a sum of two power deposition profiles. One is related to the power absorbed in a single pass, and its shape is similar to that obtained for large aspect ratio and circular cross-section. The other profile is obtained by calculating the power deposition in the limit of weak damping, in which case the wave electric field is almost constant along the cyclotron resonance layer. A heuristic formula for the power deposition is given. The formula includes a number of calibration curves and functions which has been calculated with the LION code for JET relevant equilibria. The formula enables calculation of the power deposition profile in a simple way when the launched wave spectrum and damping coefficients are known. (author). 7 refs, 11 figs
Detection of tokamak plasma positrons using annihilation photons
Energy Technology Data Exchange (ETDEWEB)
Guanying, Yu; Liu, Jian; Xie, Jinlin [University of Science and Technology, Hefei, Anhui, 230027 (China); Li, Jiangang, E-mail: j_li@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)
2017-05-15
Highlights: • A design for detection of tokamak plasma positrons is given. • Identify the main obstacle toward experimental confirmation of fusion plasma positrons. • Signal to noise ratio in a plasma disruption is estimated. • Unique potential applications of fusion plasma positrons are discussed. - Abstract: A massive amount of positrons (plasma positrons), produced by the collision between runaway electrons and nuclei during fusion plasma disruption, was first predicted theoretically in 2003. To help confirm this prediction, we report here the design of an experimental system to detect tokamak plasma positrons. Because a substantial amount of positrons (material positrons) are produced when runaway electrons impact plasma-facing materials, we proposed maximizing the ratio of plasma to material positrons by inserting a thin carbon target at the plasma edge as a plasma positron bombing target and producing a plasma disruption scenario triggered by massive gas injection. Meanwhile, the coincidence detection of positron annihilation photons was used to filter out the noise of annihilation photons from locations other than the carbon target and that of bremsstrahlung photons near 511 keV. According to our simulation, the overall signal-to-noise ratio should be more than 10:1.
Magnetic field structure near the plasma boundary in helical systems and divertor tokamaks
International Nuclear Information System (INIS)
Nagasaki, Kazunobu; Itoh, Kimitaka
1990-02-01
Magnetic field structure of the scrape off layer (SOL) region in both helical systems and divertor tokamaks is studied numerically by using model fields. The connection length of the field line to the wall is calculated. In helical systems, the connection length, L, has a logarithmic dependence on the distance from the outermost magnetic surface or that from the residual magnetic islands. The effect of axisymmetric fields on the field structure is also determined. In divertor tokamaks, the connection length also has logarithmic properties near the separatrix. Even when the perturbations, which resonate to rational surfaces near the plasma boundary, are added, logarithmic properties still remain. We compare the connection length of torsatron/helical-heliotron systems with that of divertor tokamaks. It is found that the former is shorter than the latter by one order magnitude with similar aspect ratio. (author)
The use of scaling laws for the design of high beta tokamaks
International Nuclear Information System (INIS)
Mauel, M.E.
1987-01-01
Several different empirical scaling laws for the tokamak energy confinement time are used to estimate the auxiliary heating power required for a laboratory experiment capable of testing tokamak confinement at high beta and techniques to access the second stability regime. Since operating experience in the second stability regime does not yet exist, these laws predict a wide range of possible power requirements, especially at large aspect ratios. However, by examining a model DT fusion power reactor with reasonable restrictions on the fusion island weight, neutron loading, and maximum magnetic field of the external coils, only a limited range of operating conditions are found for both first and second regime tokamaks, and only a subset of the scaling laws predict ignition. These particular scaling laws are then used to set confinement goals which if demonstrated by the laboratory experiment would indicate favourable scaling to a reactor. (author)
Serov, S. V.; Tugarinov, S. N.; Klyuchnikov, L. A.; Krupin, V. A.; von Hellermann, M.
2017-12-01
The applicability of the CXSFIT code to process experimental data from Charge-eXchange Recombination Spectroscopy (CXRS) diagnostics at the T-10 tokamak is studied with a view to its further use for processing experimental data at the ITER facility. The design and operating principle of the CXRS diagnostics are described. The main methods for processing the CXRS spectra of the 5291-Å line of C5+ ions at the T-10 tokamak (with and without subtraction of parasitic emission from the edge plasma) are analyzed. The method of averaging the CXRS spectra over several shots, which is used at the T-10 tokamak to increase the signal-to-noise ratio, is described. The approximation of the spectrum by a set of Gaussian components is used to identify the active CXRS line in the measured spectrum. Using the CXSFIT code, the ion temperature in ohmic discharges and discharges with auxiliary electron cyclotron resonance heating (ECRH) at the T-10 tokamak is calculated from the CXRS spectra of the 5291-Å line. The time behavior of the ion temperature profile in different ohmic heating modes is studied. The temperature profile dependence on the ECRH power is measured, and the dynamics of ECR removal of carbon nuclei from the T-10 plasma is described. Experimental data from the CXRS diagnostics at T-10 substantially contribute to the implementation of physical programs of studies on heat and particle transport in tokamak plasmas and investigation of geodesic acoustic mode properties.
International Nuclear Information System (INIS)
Dini, F.; Khorasani, S.
2007-01-01
Configuration of Tokamak plasma has a dominant effect on its parameters. In the calculation of transport, there are some transport coefficients and quantities, where the knowledge of their precise values, according to the system of equations, is essential to be realized. Tokamak has a toroidal configuration, in addition to classical effects, it is necessary to study the neoclassical effects due to the field curvature. The trapped particles in strong electromagnetic fields oscillate on banana-shaped orbits which in turn affect many other collisional transport parameters. Here, a precise estimation of trapped particles based on the standard equilibrium model for an elliptical shape of Tokamak plasma has been carried out using Lin-Liu model. It should be added that in this calculation, the profile of the averaged magnetic field on the flux surfaces has been derived using analytical integration and consideration of an elliptic shape for ellipticity function in the limit of large aspect ratio and zero shift of magnetic flux surfaces. Having the fraction of the trapped particles, by ,following the formulation and using an appropriate model in various collisional regimes, the neoclassical conductivity of plasma in Damavand Tokamak is obtained and the respective variations have been found. The presented results can exploit the computation of transport and other quantities of Damavand Tokamak
Optimization of magnetic field system for glass spherical tokamak GLAST-III
International Nuclear Information System (INIS)
Ahmad, Zahoor; Ahmad, S; Naveed, M A; Deeba, F; Javeed, M Aqib; Batool, S; Hussain, S; Vorobyov, G M
2017-01-01
GLAST-III (Glass Spherical Tokamak) is a spherical tokamak with aspect ratio A = 2. The mapping of its magnetic system is performed to optimize the GLAST-III tokamak for plasma initiation using a Hall probe. Magnetic field from toroidal coils shows 1/ R dependence which is typical with spherical tokamaks. Toroidal field (TF) coils can produce 875 Gauss field, an essential requirement for electron cyclotron resonance assisted discharge. The central solenoid (CS) of GLAST-III is an air core solenoid and requires compensation coils to reduce unnecessary magnetic flux inside the vessel region. The vertical component of magnetic field from the CS in the vacuum vessel region is reduced to 1.15 Gauss kA −1 with the help of a differential loop. The CS of GLAST can produce flux change up to 68 mVs. Theoretical and experimental results are compared for the current waveform of TF coils using a combination of fast and slow capacitor banks. Also the magnetic field produced by poloidal field (PF) coils is compared with theoretically predicted values. It is found that calculated results are in good agreement with experimental measurement. Consequently magnetic field measurements are validated. A tokamak discharge with 2 kA plasma current and pulse length 1 ms is successfully produced using different sets of coils. (paper)
West, Stuart A; Reece, S E; Sheldon, Ben C
2002-01-01
Sex ratio theory attempts to explain variation at all levels (species, population, individual, brood) in the proportion of offspring that are male (the sex ratio). In many cases this work has been extremely successful, providing qualitative and even quantitative explanations of sex ratio variation. However, this is not always the situation, and one of the greatest remaining problems is explaining broad taxonomic patterns. Specifically, why do different organisms show so ...
Plasma position control in TCABR Tokamak
International Nuclear Information System (INIS)
Galvao, R.M.O.; Kuznetsov, Yu. K.; Nascimento, I.C.; Fonseca, A.M.M.; Silva, R.P. da; Ruchko, L.F.; Tuszel, A.G.; Reis, A.P. dos; Sanada, E.K.
1998-01-01
The plasma control position in the TCABR tokamak is described. The TCA tokamak was transferred from the Centre de Recherches en Physique des Plasmas, Lausanne, to the Institute of Physics of University of Sao Paulo, renamed TCABR (α=0.18 m, R = 0.62 m, B = 1 T,I p = 100 kA). The control system was reconstructed using mainly components obtained from the TCA tokamak. A new method of plasma position determination is used in TCABR to improve its accuracy. A more detailed theoretical analysis of the feed forward and feedback control is performed as compared with. (author)
Estimation of Zeff in Novillo Tokamak
International Nuclear Information System (INIS)
Valencia, R.; Olayo, G.; Cruz, G.; Lopez, R.; Chavez, E.; Melendez, L.; Flores, A.; Gaytan, E.
1996-01-01
We estimated the Z eff in the Novillo Tokamak after having applied a HeGDC process through two different methods: anomaly factor and mass spectrometry. The first one gave a Z eff value of 2.07 for a tokamak discharge of 4350 A plasma current and 3 V of loop voltage. By mass spectrometry 30 s after the discharge had finished a Z eff of 4.19 was obtained for the same discharge. By mass spectrometry we observed that the Z eff value is a time function. Furthermore this method is helpful for evaluating the level of impurities after many discharges in Novillo Tokamak. (orig.)
Global gyrokinetic simulation of tokamak transport
International Nuclear Information System (INIS)
Furnish, G.; Horton, W.; Kishimoto, Y.; LeBrun, M.J.; Tajima, T.
1998-10-01
A kinetic simulation code based on the gyrokinetic ion dynamics in global general metric (including a tokamak with circular or noncircular cross-section) has been developed. This gyrokinetic simulation is capable of examining the global and semi-global driftwave structures and their associated transport in a tokamak plasma. The authors investigate the property of the ion temperature gradient (ITG) or η i (η i ≡ ∂ ell nT i /∂ ell n n i ) driven drift waves in a tokamak plasma. The emergent semi-global drift wave modes give rise to thermal transport characterized by the Bohm scaling
Fast IR diodes thermometer for tokamak
International Nuclear Information System (INIS)
Chen Xiangbo
2001-01-01
A 30 channel fast IR pyrometry array has been constructed for tokamak, which has 0.5 μs time response, 10 mm diameter spatial resolution and 5 degree C temperature resolution. The temperature measuring range is from 250 degree C to 1200 degree C. The two dimensional temperature profiles of the first wall during both major and minor disruptions can be measured with an accuracy of about 1% measuring temperature, which is adequate for tokamak experiments. This gives a very useful tool for the disruption study, especially for the divertor physics and edge heat flux research on tokamak and other magnetic confinement devices
Linear neoclassical tearing mode in tokamaks
International Nuclear Information System (INIS)
Shaing, K. C.
2007-01-01
The growth rate of linear tearing modes in tokamaks is calculated including the neoclassical dissipation mechanism. It is found that when the growth rate is much smaller than the ion-ion collision frequency, the growth rate is reduced approximately by a factor of (B p /B) 2/5 from its standard value, and when the growth rate is much larger than the ion-ion collision frequency, the growth rate is reduced by a factor [√(ε)/(1.6q 2 )] 1/5 . Here, B p is the poloidal magnetic field strength, B is the magnetic field strength, ε is the inverse aspect ratio, and q is the safety factor. The width of the resistive layer is broadened when compared to that of the standard theory. In both limits, the growth rate and the resistive layer width only depend on B p and are independent of B. The growth rates in the plateau regime and for the inertia dominant modes are also presented
Research and development of JT-60 tokamak
International Nuclear Information System (INIS)
Saito, Ryusei; Sato, Hiroshi; Murata, Toshifumi; Ito, Yoshiyasu.
1978-01-01
The development of nuclear fusion apparatuses for the purpose of utilizing energy due to nuclear fusion reaction has been forwarded in various countries, and in Japan, the critical plasma testing apparatus JT-60 is about to be constructed, centering around Japan Atomic Energy Research Institute. This is one of four large apparatuses to be constructed in the world, and it is expected to be completed in 1982. JT-60 is a nuclear fusion apparatus of tokamak type aiming at generating critical plasma. The features of JT-60 are the formation of the plasma with small aspect ratio, the equipment of a magnetic limiter, the arrangement of the first wall of molybdenum and high temperature baking as the measures to impurities. The large toroidal magnetic field coil of JT-60 is composed of 18 unit coils. The analyses of magnetic field, thermal behavior and structural strength, the selection of materials, and the development of manufacturing techniques regarding the toroidal coil are described. The vacuum container of JT-60 is composed of the main body of torus type comprising thickwalled rings and bellows, the first wall comprising liners, fixed limiter and magnetic limiter, and observation ports. It is large torus-form container with non-circular cross section, and baking at 500 deg. C is required as the measure to ultrahigh vacuum. Complex forces including electromagnetic force act on it. (Kako, I.)
Multiple time scale methods in tokamak magnetohydrodynamics
International Nuclear Information System (INIS)
Jardin, S.C.
1984-01-01
Several methods are discussed for integrating the magnetohydrodynamic (MHD) equations in tokamak systems on other than the fastest time scale. The dynamical grid method for simulating ideal MHD instabilities utilizes a natural nonorthogonal time-dependent coordinate transformation based on the magnetic field lines. The coordinate transformation is chosen to be free of the fast time scale motion itself, and to yield a relatively simple scalar equation for the total pressure, P = p + B 2 /2μ 0 , which can be integrated implicitly to average over the fast time scale oscillations. Two methods are described for the resistive time scale. The zero-mass method uses a reduced set of two-fluid transport equations obtained by expanding in the inverse magnetic Reynolds number, and in the small ratio of perpendicular to parallel mobilities and thermal conductivities. The momentum equation becomes a constraint equation that forces the pressure and magnetic fields and currents to remain in force balance equilibrium as they evolve. The large mass method artificially scales up the ion mass and viscosity, thereby reducing the severe time scale disparity between wavelike and diffusionlike phenomena, but not changing the resistive time scale behavior. Other methods addressing the intermediate time scales are discussed
Tokamak power system studies at ANL
International Nuclear Information System (INIS)
Baker, C.C.; Ehst, D.A.; Brooks, J.N.; Evans, K. Jr.
1986-06-01
The following features, in particular, have been examined: (a) large aspect ratio (A ≅ 6), which may ease maintenance; (b) high beta (β ≥ 0.20) without indentation, which brings the maximum toroidal field down to about 6 to 7 T; (c) low toroidal current (I ≅ 4MA), which reduces the cost of the current drive and equilibrium field system; and (d) steady state operation with current density control via fast and slow wave current drive. The key to high beta operation with low toroidal current lies in utilizing second stability regime equilibria with the required current distributions produced by an appropriate selection of wave driver frequencies and power spectra. The ray tracing and current drive calculation is self-consistent with the actual magnetic fields they produce in the plasma. The impurity control activities in TPSS have emphasized the self-pumping concept as applied to using the entire first wall or ''slot'' limiters. The blanket design effort has emphasized liquid metal and Flibe concepts. The reference concept is a liquid lithium/vanadium, self-cooled configuration. Overall, there exists a number of major design improvements which will substantially improve the attractiveness of tokamak reactors
Vdovin, V.
2014-02-01
The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].
Energy Technology Data Exchange (ETDEWEB)
Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)
2014-02-12
The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].
Submillimeter wave propagation in tokamak plasmas
International Nuclear Information System (INIS)
Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.
1985-01-01
The propagation of submillimeter-waves (smm) in tokamak plasmas has been investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses have been carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system has been employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes have been developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements
Power and particle exhaust in tokamaks
International Nuclear Information System (INIS)
Stambaugh, R.D.
1998-01-01
The status of power and particle exhaust research in tokamaks is reviewed in the light of ITER requirements. There is a sound basis for ITER's nominal design positions; important directions for further research are identified
Definition of total bootstrap current in tokamaks
International Nuclear Information System (INIS)
Ross, D.W.
1995-01-01
Alternative definitions of the total bootstrap current are compared. An analogous comparison is given for the ohmic and auxiliary currents. It is argued that different definitions than those usually employed lead to simpler analyses of tokamak operating scenarios
Tokamak research in the Soviet Union
International Nuclear Information System (INIS)
Strelkov, V.S.
1981-01-01
Important milestones on the way to the tokamak fusion reactor are recapitulated. Soviet tokamak research concentrated at the I.V. Kurchatov Institute in Moscow, the A.F. Ioffe Institute in Leningrad and the Physical-Technical Institute in Sukhumi successfully provides necessary scientific and technological data for reactor design. Achievments include, the successful operation of the first tokamak with superconducting windings (T-7) and the gyrotron set for microwave plasma heating in the T-10 tokamak. The following problems have intensively been studied: Various methods of additional plasma heating, heat and particle transport, and impurity control. The efficiency of electron-cyclotron resonance heating was demonstrated. In the Joule heating regime, both the heat conduction and diffusion rates are anomalously high, but the electron heat conduction rate decreases with increasing plasma density. Progress in impurity control makes it possible to obtain a plasma with effective charge approaching unity. (J.U.)
Neutral beam in ALVAND IIC tokamak
International Nuclear Information System (INIS)
Ghrannevisse, M.; Moradshahi, M.; Avakian, M.
1992-01-01
Neutral beams have a wide application in tokamak experiments. It used to heat; fuel; adjust electric potentials in plasmas and diagnose particles densities and momentum distributions. It may be used to sustain currents in tokamaks to extend the pulse length. A 5 KV; 500 mA ion source has been constructed by plasma physics group, AEOI and it used to produce plasma and study the plasma parameters. Recently this ion source has been neutralized and it adapted to a neutral beam source; and it used to heat a cylindrical DC plasma and the plasma of ALVAND IIC Tokamak which is a small research tokamak with a minor radius of 12.6 cm, and a major radius of 45.5 cm. In this paper we report the neutralization of the ion beam and the results obtained by injection of this neutral beam into plasmas. (author) 2 refs., 4 figs
Submillimeter wave propagation in tokamak plasmas
International Nuclear Information System (INIS)
Ma, C.H.; Hutchinson, D.P.; Staats, P.A.; Vander Sluis, K.L.; Mansfield, D.K.; Park, H.; Johnson, L.C.
1986-01-01
Propagation of submillimeter waves (smm) in tokamak plasma was investigated both theoretically and experimentally to ensure successful measurements of electron density and plasma current distributions in tokamak devices. Theoretical analyses were carried out to study the polarization of the smm waves in TFTR and ISX-B tokamaks. A multichord smm wave interferometer/polarimeter system was employed to simultaneously measure the line electron density and poloidal field-induced Faraday rotation in the ISX-B tokamak. The experimental study on TFTR is under way. Computer codes were developed and have been used to study the wave propagation and to reconstruct the distributions of plasma current and density from the measured data. The results are compared with other measurements. 5 references, 2 figures
Empirical scaling for present Ohmically heated tokamaks
International Nuclear Information System (INIS)
Daughney, C.
1975-01-01
Experimental results from the Adiabatic Toroidal Compressor (ATC) tokamak are used to obtain empirical scaling laws for the average electron temperature and electron energy confinement time as functions of the average electron density, the effective ion charge, and the plasma current. These scaling laws are extended to include dependence upon minor and major plasma radius and toroidal field strength through a comparison of the various tokamaks described in the literature. Electron thermal conductivity is the dominant loss process for the ATC tokamak. The parametric dependences of the observed electron thermal conductivity are not explained by present theoretical considerations. The electron temperature obtained with Ohmic heating is shown to be a function of current density - which will not be increased in the next generation of large tokamaks. However, the temperature dependence of the electron energy confinement time suggests that significant improvement in confinement time will be obtained with supplementary electron heating. (author)
Research into controlled fusion in tokamaks
International Nuclear Information System (INIS)
Zacek, F.
1992-01-01
During the thirty years of tokamak research, physicists have been approaching step by step the reactor breakeven condition defined by the Lawson criterion. JET, the European Community tokamak is probably the first candidate among the world largest tokamaks to reach the ignition threshold and thus to demonstrate the physical feasibility of thermonuclear reaction. The record plasma parameters achieved in JET at H plasma modes due to powerful additional plasma heating and due to substantial reduction of plasma impurities, opened the door to the first experiment with a deuterium-tritium plasma. In the paper, the conditions and results of these tritium experiments are described in detail. The prospects of the world tokamak research and of the participation of Czechoslovak physicists are also discussed. (J.U.) 3 figs., 6 refs
Robust Sliding Mode Control for Tokamaks
Directory of Open Access Journals (Sweden)
I. Garrido
2012-01-01
Full Text Available Nuclear fusion has arisen as an alternative energy to avoid carbon dioxide emissions, being the tokamak a promising nuclear fusion reactor that uses a magnetic field to confine plasma in the shape of a torus. However, different kinds of magnetohydrodynamic instabilities may affect tokamak plasma equilibrium, causing severe reduction of particle confinement and leading to plasma disruptions. In this sense, numerous efforts and resources have been devoted to seeking solutions for the different plasma control problems so as to avoid energy confinement time decrements in these devices. In particular, since the growth rate of the vertical instability increases with the internal inductance, lowering the internal inductance is a fundamental issue to address for the elongated plasmas employed within the advanced tokamaks currently under development. In this sense, this paper introduces a lumped parameter numerical model of the tokamak in order to design a novel robust sliding mode controller for the internal inductance using the transformer primary coil as actuator.
Plasma equilibrium and instabilities in tokamaks
International Nuclear Information System (INIS)
Caldas, I.L.; Vannucci, A.
1985-01-01
A phenomenological introduction of some of the main theoretical and experimental features on equilibrium and instabilities in tokamaks is presented. In general only macroscopic effects are considered, being the plasma described as a fluid. (L.C.) [pt
The ARIES-I tokamak reactor study
International Nuclear Information System (INIS)
1991-01-01
This report discusses the following topics on the Aries-I Tokamak: Design description; systems studies and economics; reactor plasma physics; magnet engineering; fusion-power-ore engineering; and environmental and safety features
Theory of incremental turbulent transport in tokamaks
International Nuclear Information System (INIS)
Similon, P.L.
1991-01-01
The goal of this research is to understand how the various aspect of turbulent transport operate in tokamaks, in the presence of low frequency fluctuations such as drift waves or trapped electron modes
Numerical simulation of edge plasma in tokamak
International Nuclear Information System (INIS)
Chen Yiping; Qiu Lijian
1996-02-01
The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)
Plasma diagnostics using synchrotron radiation in tokamaks
International Nuclear Information System (INIS)
Fidone, I.; Giruzzi, G.; Granata, G.
1995-09-01
This report deal with the use of synchrotron radiation in tokamaks. The main advantage of this new method is that it enables to overcome several deficiencies, caused by cut-off, refraction, and harmonic overlap. It also makes it possible to enhance the informative contents of the familiar low harmonic scheme. The basic theory of the method is presented and illustrated by numerical applications, for plasma parameters of relevance in present and next step tokamaks. (TEC). 10 refs., 13 figs
The ETE spherical Tokamak project. IAEA report
Energy Technology Data Exchange (ETDEWEB)
Ludwig, Gerson Otto; Del Bosco, E.; Berni, L.A.; Ferreira, J.G.; Oliveira, R.M.; Andrade, M.C.R.; Shibata, C.S.; Barroso, J.J.; Castro, P.J.; Patire Junior, H. [Instituto Nacional de Pesquisas Espaciais (INPE), Sao Jose dos Campos, SP (Brazil). Lab. Associado de Plasma]. E-mail: ludwig@plasma.inpe.br
2002-07-01
This paper describes the general characteristics of spherical tokamaks, or spherical tori, with a brief overview of work in this area already performed or in progress at several institutions worldwide. The paper presents also the historical development of the ETE (Spherical Tokamak Experiment) project, its research program, technical characteristics and operating conditions as of October, 2002 at the Associated Plasma Laboratory (LAP) of the National Space Research Institute (INPE) in Brazil. (author)
Electric conductivity and bootstrap current in tokamak
International Nuclear Information System (INIS)
Mao Jianshan; Wang Maoquan
1996-12-01
A modified Ohm's law for the electric conductivity calculation is presented, where the modified ohmic current can be compensated by the bootstrap current. A comparison of TEXT tokamak experiment with the theories shows that the modified Ohm's law is a more close approximation to the tokamak experiments than the classical and neoclassical theories and can not lead to the absurd result of Z eff <1, and the extended neoclassical theory would be not necessary. (3 figs.)
Erosion of the first wall of Tokamaks
International Nuclear Information System (INIS)
Guseva, M.I.; Ionova, E.S.; Martynenko, Yu.V.
1980-01-01
An estimate of the rate of erosion of the wall due to sputtering and blistering requires knowledge of the fluxes and energies of the particles which go from the plasma to the wall, of the sputtering coefficients S, and of the erosion coefficients S* for blistering. The overall erosion coefficient is equal to the sum of the sputtering coefficient and the erosion coefficient for blistering. Here the T-20 Tokamak is examined as an example of a large-scale Tokamak. 18 refs
Advanced tokamak physics in DIII-D
Energy Technology Data Exchange (ETDEWEB)
Petty, C.C.; Luce, T.C.; Politzer, P.A.; Bray, B.; Burrell, K.H.; Chu, M.S.; Ferron, J.R.; Gohil, P.; Greenfield, C.M.; Hsieh, C.-L.; Hyatt, A.W.; La Haye, R.J.; Lao, L.L.; Leonard, A.W.; Lin-Liu, Y.R.; Lohr, J.; Mahdavi, M.A.; Petrie, T.W.; Pinsker, R.I.; Prater, R.; Scoville, J.T.; Staebler, G.M.; Strait, E.J.; Taylor, T.S.; West, W.P. [General Atomics, PO Box 85608, San Diego, CA (United States); Wade, M.R.; Lazarus, E.A.; Murakami, M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Allen, S.L.; Casper, T.A.; Jayakumar, R.; Lasnier, C.J.; Makowski, M.A.; Rice, B.W.; Wolf, N.S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Austin, M.E. [University of Texas, Austin, TX (United States); Fredrickson, E.D.; Gorelov, I.; Johnson, L.C.; Okabayashi, M.; Wong, K.-L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Garofalo, A.M.; Navratil, G.A. [Columbia University, New York (United States); Heidbrink, W. [University of California, Irvine, CA (United States); Kinsey, J.E. [Leheigh University, Bethlehem, PA (United States); McKee, G.R. [University of Wisconsin, Madison, WI (United States); Rettig, C.L.; Rhodes, T.L. [University of California, Los Angeles, CA (United States); Watkins, J.G. [Sandia National Laboratories, Albuquerque, NM (United States)
2000-12-01
Advanced tokamaks seek to achieve a high bootstrap current fraction without sacrificing fusion power density or fusion gain. Good progress has been made towards the DIII-D research goal of demonstrating a high-{beta} advanced tokamak plasma in steady state with a relaxed, fully non-inductive current profile and a bootstrap current fraction greater than 50%. The limiting factors for transport, stability, and current profile control in advanced operating modes are discussed in this paper. (author)
Optimization design for SST-1 Tokamak insulators
International Nuclear Information System (INIS)
Zhang Yuanbin; Pan Wanjiang
2012-01-01
With the help of ANSYS FEA technique, high voltage and cryogenic proper- ties of the SST-1 Tokamak insulators were obtained, and the structure of the insulators was designed and modified by taking into account the simulation results. The simulation results indicate that the optimization structure has better high voltage insulating property and cryogenic mechanics property, and also can fulfill the qualification criteria of the SST-1 Tokamak insulators. (authors)
Spherical tokamak without external toroidal fields
International Nuclear Information System (INIS)
Kaw, P.K.; Avinash, K.; Srinivasan, R.
2001-01-01
A spherical tokamak design without external toroidal field coils is proposed. The tokamak is surrounded by a spheromak shell carrying requisite force free currents to produce the toroidal field in the core. Such equilibria are constructed and it is indicated that these equilibria are likely to have robust ideal and resistive stability. The advantage of this scheme in terms of a reduced ohmic dissipation is pointed out. (author)
Preliminary results of the TBR small tokamak
International Nuclear Information System (INIS)
Nascimento, I.C.; Fagundes, A.N.; Da Silva, R.P.; Galvao, R.M.O.; Del Bosco, E.; Vuolo, J.H.; Sanada, E.K.; Dellaqua, R.
1982-01-01
The paper gives a short description of the TBR - small Brazilian tokamak and the first results obtained for plasma formation and equilibrium. Measured breakdown curves for hydrogen are shown to be confined within analytically calculated limits and to depend strongly on stray vertical magnetic fields. Time profiles of plasma current in equilibrium are shown and compared with the predictions of a simple analytical model for tokamak discharges. Reasonable agreement is obtained taking Zsub(eff) as a free parameter. (author)
Full power in the European tokamak
International Nuclear Information System (INIS)
Lallia, P.P.; Hugon, M.
1987-01-01
A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years
Facility approach to tokamak operation
International Nuclear Information System (INIS)
Edmonds, P.H.; Gabbard, W.A.
1981-01-01
In anticipation of the appearance of more advanced tokamaks and other fusion relevant experiments, program has been established at ORNL to systemically identify the requirements of an effective machine operations group. This program is presently applied to the ISX-B experiment. With its continuing development, it is expected to provide major support in the identification of potential problem areas and to assist in the generation of the necessary procedures for forthcoming devices. The present and future generations of large plasma devices will function as facilities, operated by an operations group as service to the plasma physicists and diagnosticians. The purpose of the program discussed here is to develop and to encourage an orderly transition to the facility-like style of operation
Developments in tokamak transport modeling
International Nuclear Information System (INIS)
Houlberg, W.A.; Attenberger; Lao, L.L.
1981-01-01
A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements
Simulation models for tokamak plasmas
International Nuclear Information System (INIS)
Dimits, A.M.; Cohen, B.I.
1992-01-01
Two developments in the nonlinear simulation of tokamak plasmas are described: (A) Simulation algorithms that use quasiballooning coordinates have been implemented in a 3D fluid code and a 3D partially linearized (Δf) particle code. In quasiballooning coordinates, one of the coordinate directions is closely aligned with that of the magnetic field, allowing both optimal use of the grid resolution for structures highly elongated along the magnetic field as well as implementation of the correct periodicity conditions with no discontinuities in the toroidal direction. (B) Progress on the implementation of a likeparticle collision operator suitable for use in partially linearized particle codes is reported. The binary collision approach is shown to be unusable for this purpose. The algorithm under development is a complete version of the test-particle plus source-field approach that was suggested and partially implemented by Xu and Rosenbluth
Burn Control Mechanisms in Tokamaks
Hill, M. A.; Stacey, W. M.
2015-11-01
Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.
Tokamak plasma interaction with limiters
International Nuclear Information System (INIS)
Pitcher, C.S.
1987-11-01
The importance of plasma purity is first discussed in terms of the general requirements of controlled thermonuclear fusion. The tokamak approach to fusion and its inherent problem of plasma contamination are introduced. A main source of impurities is due to the bombardment of the limiter by energetic particles and thus the three main aspects of the plasma-limiter interaction are reviewed, boundary plasma conditions, fuelling/recycling and impurity production. The experiments, carried out on the DITE tokamak at Culham Laboratory, UK, investigated these three topics and the results are compared with predicted behaviour; new physical phenomena are presented in all three areas. Simple one-dimensional fluid equations are found to adequately describe the SOL plasma, except in regard to the pre-sheath electric field and ambipolarity; that is, the electric field adjacent to the limiter surface appears to be weak and the associated plasma flow can be non-ambipolar. Recycling of fuel particles from the limiter is observed to be near unity at all times. The break-up behaviour of recycled and gas puffed D 2 molecules is dependent on the electron temperature, as expected. Impurity production at the limiter is chemical erosion of graphite being negligible. Deposition of limiter and wall-produced impurities is found on the limiter. The spatial distributions of impurities released from the limiter are observed and are in good agreement with a sputtered atom transport code. Finally, preliminary experiments on the transport of impurity ions along field lines away from the limiter have been performed and compared with simple analytic theory. The results suggest that the pre-sheath electric field in the SOL is much weaker than the simple fluid model would predict
Economic considerations of commercial tokamak options
International Nuclear Information System (INIS)
Dabiri, A.E.
1986-05-01
Systems studies have been performed to assess commercial tokamak options. Superconducting, as well as normal, magnet coils in either first or second stability regimes have been considered. A spherical torus (ST), as well as an elongated tokamak (ET), is included in the study. The cost of electricity (COE) is selected as the figure of merit, and beta and first-wall neutron wall loads are selected to represent the physics and technology characteristics of various options. The results indicate that an economical optimum for tokamaks is predicted to require a beta of around 10%, as predicted to be achieved in the second stability regime, and a wall load of about 5 MW/m 2 , which is assumed to be optimum technologically. This tokamak is expected to be competitive with fission plants if efficient, noninductive current drive is developed. However, if this regime cannot be attained, all other tokamaks operating in the first stability regime, including spherical torus and elongated tokamak and assuming a limiting wall load of 5 MW/m 2 , will compete with one another with a COE of about 50 mill/kWh. This 40% higher than the COE for the optimum reactor in the second stability regime with fast-wave current drive. The above conclusions pertain to a 1200-MW(e) net electric power plant. A comparison was also made between ST, ET, and superconducting magnets in the second stability regime with fast-wave current drive at 600 MW(e)
D-D tokamak reactor assessment
International Nuclear Information System (INIS)
Baxter, D.C.; Dabiri, A.E.
1983-01-01
A quantitative comparison of the physics and technology requirements, and the cost and safety performance of a d-d tokamak relative to a d-t tokamak has been performed. The first wall/blanket and energy recovery cycle for the d-d tokamak is simpler, and has a higher efficiency than the d-t tokamak. In most other technology areas (such as magnets, RF, vacuum, etc.) d-d requirements are more severe and the systems are more complex, expensive and may involve higher technical risk than d-t tokamak systems. Tritium technology for processing the plasma exhaust, and tritium refueling technology are required for d-d reactors, but no tritium containment around the blanket or heat transport system is needed. Cost studies show that for high plasma beta and high magnetic field the cost of electricity from d-d and d-t tokamaks is comparable. Safety analysis shows less radioactivity in a d-d reactor but larger amounts of stored energy and thus higher potential for energy release. Consequences of all postulated d-d accidents are significantly smaller than those from d-t reactor tritium releases
Measurement of the effective plasma ion mass in large tokamaks
International Nuclear Information System (INIS)
Lister, J.B.; Villard, L.; Ridder, G. de
1997-01-01
There is not yet a straightforward method for the measurement of the D-T ratio in the centre of a tokamak plasma. One of the simpler measurements put forward in the past is the interpretation of the MHD spectrum in the frequency range of the Global Alfven Eigenmodes (GAE). However, the frequencies of these modes do not only depend on the plasma mass, but are also quite strongly dependent on the details of the current and density profiles, creating a problem of deconvolution of the estimate of the plasma mass from an implicit relationship between several measurable plasma parameters and the detected eigenmode frequencies. This method has been revised to assess its likely precision for the JET tokamak. The low n GAE modes are sometimes too close to the continuum edge to be detectable and the interpretation of the GAE spectrum is rendered less direct than had been hoped. We present a statistical study on the precision with which the D-T ratio could be estimated from the GAE spectrum on JET. (author) 4 figs., 8 refs
International Nuclear Information System (INIS)
Romanelli, M.
1999-09-01
The poloidal variation of impurity densities over magnetic surfaces brings about an enhancement of neoclassical transport coefficients, as shown by Romanelli and Ottaviani for impurities in the Pfirsch Schlueter regime and by Helander for particles in the banana-plateau regime, both in a large aspect ratio tokamak. The same effect will occur in a finite aspect ratio tokamak and therefore it is considered to be relevant for inclusion in transport codes for comparison with the experimental measurements of impurity transport. Here an expression for the impurity-density poloidal-variation generated by the fast toroidal rotation of the plasma column is presented in general coordinates. (author)
Stabe, R. G.; Whitney, W. J.; Moffitt, T. P.
1984-01-01
Experimental results are presented for a 0.767 scale model of the first stage of a two-stage turbine designed for a high by-pass ratio engine. The turbine was tested with both uniform inlet conditions and with an inlet radial temperature profile simulating engine conditions. The inlet temperature profile was essentially mixed-out in the rotor. There was also substantial underturning of the exit flow at the mean diameter. Both of these effects were attributed to strong secondary flows in the rotor blading. There were no significant differences in the stage performance with either inlet condition when differences in tip clearance were considered. Performance was very close to design intent in both cases.
Implications of rf current drive theory for next step steady-state tokamak design
International Nuclear Information System (INIS)
Schultz, J.H.
1985-06-01
Two missions have been identified for a next-step tokamak experiment in the United States. The more ambitious Mission II device would be a superconducting tokamak, capable of doing long-pulse ignition demonstrations, and hopefully capable of also being able to achieve steady-state burn. A few interesting lines of approach have been identified, using a combination of logical design criteria and parametric system scans [SC85]. These include: (1) TIBER: A point-design suggested by Lawrence Livermore, that proposes a machine with the capability of demonstrating ignition, high beta (10%) and high Q (=10), using high frequency, fast-wave current drive. The TIBER topology uses moderate aspect ratio and high triangularity to achieve high beta. (2) JET Scale-up. (3) Magic5: It is argued here that an aspect ratio of 5 is a magic number for a good steady-state current drive experiment. A moderately-sized machine that achieves ignition and is capable of high Q, using either fast wave or slow wave current drive is described. (4) ET-II: The concept of a highly elongated tokamak (ET) was first proposed as a low-cost approach to Mission I, because of the possibility of achieving ohmic ignition with low-stress copper magnets. We propose that its best application is really for commercial tokamaks, using fast-wave current drive, and suggest a Mission II experiment that would be prototypical of such a reactor
About the Toroidal Magnetic Field of a Tokamak Burning Plasma Experiment with Superconducting Coils
International Nuclear Information System (INIS)
Mazzucato, E.
2002-01-01
In tokamaks, the strong dependence on the toroidal magnetic field of both plasma pressure and energy confinement is what makes possible the construction of small and relatively inexpensive burning plasma experiments using high-field resistive coils. On the other hand, the toroidal magnetic field of tokamaks using superconducting coils is limited by the critical field of superconductivity. In this article, we examine the relative merit of raising the magnetic field of a tokamak plasma by increasing its aspect ratio at a constant value of the peak field in the toroidal magnet. Taking ITER-FEAT as an example, we find that it is possible to reach thermonuclear ignition using an aspect ratio of approximately 4.5 and a toroidal magnetic field of 7.3 T. Under these conditions, fusion power density and neutron wall loading are the same as in ITER [International Thermonuclear Experimental Reactor], but the normalized plasma beta is substantially smaller. Furthermore, such a tokamak would be able to reach an energy gain of approximately 15 even with the deterioration in plasma confinement that is known to occur near the density limit where ITER is forced to operate
Disassembly of JT-60 tokamak device and ancillary facilities for JT-60 tokamak
International Nuclear Information System (INIS)
Okano, Fuminori; Ichige, Hisashi; Miyo, Yasuhiko; Kaminaga, Atsushi; Sasajima, Tadayuki; Nishiyama, Tomokazu; Yagyu, Jun-ichi; Ishige, Youichi; Suzuki, Hiroaki; Komuro, Kenichi; Sakasai, Akira; Ikeda, Yoshitaka
2014-03-01
The disassembly of JT-60 tokamak device and its peripheral equipments, where the total weight was about 5400 tons, started in 2009 and accomplished in October 2012. This disassembly was required process for JT-60SA project, which is the Satellite Tokamak project under Japan-EU international corroboration to modify the JT-60 to the superconducting tokamak. This work was the first experience of disassembling a large radioactive fusion device based on Radiation Hazard Prevention Act in Japan. The cutting was one of the main problems in this disassembly, such as to cut the welded parts together with toroidal field coils, and to cut the vacuum vessel into two. After solving these problems, the disassembly completed without disaster and accident. This report presents the outline of the JT-60 disassembly, especially tokamak device and ancillary facilities for tokamak device. (author)
Modelling ohmic confinement experiments on the START tokamak
International Nuclear Information System (INIS)
Roach, C.M.
1996-05-01
Ohmic confinement data from the tight aspect ratio tokamak START has been analysed using the ASTRA transport simulation code. Neoclassical expressions have been modified to describe tight aspect ratio configurations, and the comparison between START data and models of anomalous transport has been made quantitative using the standard χ 2 test from statistics. Four confinement models (T11, Rebut-Lallia-Watkins, Lackner-Gottardi, and Taroni et al's Bohm model) have been compared with the START data. Three of the models are found to simulate START's electron temperature data moderately well, while Taroni et al's Bohm model overestimates electron temperatures in START by an order of magnitude. Thus comparison with START data tends to discriminate against Bohm models; these models are pessimistic or ITER. (author)
Millimeter wave scattering off a whistler wave in a tokamak
International Nuclear Information System (INIS)
Sawhney, B.K.; Singh, S.V.; Tripathi, V.K.
1994-01-01
Obliquely propagating whistler waves through a plasma cause density perturbations. A high frequency electromagnetic wave sent into such a perturbed region suffers scattering. The process can be used as a diagnostics for whistler. We have developed a theory of electromagnetic wave scattering in a tokamak where density profile is taken a parabolic. Numerical calculations have been carried out to evaluate the ratio of the power of the scattered electromagnetic wave to that of the incident electromagnetic wave. The scattered power decreases with the frequency of the incident electromagnetic wave. For typical parameters, the ratio of the power of the scattered to the incident electromagnetic wave comes out to be of the order of 10 -4 at a scattering angle of 3 which can be detected. (author). 2 refs, 1 fig
On the HL-1M tokamak plasma confinement time
International Nuclear Information System (INIS)
Qin Yunwen
2001-01-01
Emphasizing that the tokamak plasma confinement time is the plasma particle or thermal energy loss characteristic time, the relevant physical concept and HL-1M tokamak experimental data analyses are reviewed
Power reduction and the radial limit of stall delay in revolving wings of different aspect ratio
Kruyt, J.W.; Heijst, Van G.F.; Altshuler, D.L.; Lentink, David
2015-01-01
Airplanes and helicopters use high aspect ratio wings to reduce the power required to fly, but must operate at low angle of attack to prevent flow separation and stall. Animals capable of slow sustained flight, such as hummingbirds, have low aspect ratio wings and flap their wings at high angle
Analysis of tokamak plasma confinement modes using the fast
Indian Academy of Sciences (India)
The Fourier analysis is a satisfactory technique for detecting plasma confinement modes in tokamaks. The confinement mode of tokamak plasma was analysed using the fast Fourier transformation (FFT). For this purpose, we used the data of Mirnov coils that is one of the identifying tools in the IR-T1 tokamak, with and ...
Comparative studies of stellarator and tokamak transport
Energy Technology Data Exchange (ETDEWEB)
Stroth, U; Burhenn, R; Geiger, J; Giannone, L.; Hartfuss, H J; Kuehner, G; Ledl, L; Simmet, E E; Walter, H [Max-Planck-Inst. fuer Plasmaphysik, IPP-Euratom Association, Garching (Germany); ECRH Team; W7-AS Team
1997-09-01
Transport properties in the W7-AS stellarator and in tokamaks are compared. The parameter dependences and the absolute values of the energy confinement time are similar. Indications are found that the density dependence, which is usually observed in stellarator confinement, can vanish above a critical density. The density dependence in stellarators seems to be similar to that in the linear ohmic confinement regime, which, in small tokamaks, extends to high density values, too. Because of the similarity in the gross confinement properties, transport in stellarators and tokamaks should not be dominated by the parameters which are very different in the two concepts, i.e. magnetic shear, major rational values of the rotational transform and plasma current. A difference in confinement is that there exists evidence for pinches in the particle and, possibly, energy transport channels in tokamaks whereas in stellarators no pinches have been observed, so far. In order to study the effect of plasma current and toroidal electric fields, stellarator discharges were carried out with an increasing amount of plasma current. From these experiments, no clear evidence of a connection of pinches with these parameters is found. The transient response in W7-AS plasmas can be described in terms of a non-local model. As in tokamaks, also cold pulse experiments in W7-AS indicate the importance of non-local transport. (author). 8 refs, 5 figs.
Three novel tokamak plasma regimes in TFTR
International Nuclear Information System (INIS)
Furth, H.P.
1985-10-01
Aside from extending ''standard'' ohmic and neutral beam heating studies to advanced plasma parameters, TFTR has encountered a number of special plasma regimes that have the potential to shed new light on the physics of tokamak confinement and the optimal design of future D-T facilities: (1) High-powered, neutral beam heating at low plasma densities can maintain a highly reactive hot-ion population (with quasi-steady-state beam fueling and current drive) in a tokamak configuration of modest bulk-plasma confinement requirements. (2) Plasma displacement away from limiter contact lends itself to clarification of the role of edge-plasma recycling and radiation cooling within the overall pattern of tokamak heat flow. (3) Noncentral auxiliary heating (with a ''hollow'' power-deposition profile) should serve to raise the central tokamak plasma temperature without deterioration of central region confinement, thus facilitating the study of alpha-heating effects in TFTR. The experimental results of regime (3) support the theory that tokamak profile consistency is related to resistive kink stability and that the global energy confinement time is determined by transport properties of the plasma edge region
Edge plasma diagnostics on Tore Supra tokamak
International Nuclear Information System (INIS)
Fujita, Junji
1991-01-01
From 1988 to 1991, the international scientific research 'Diagnosis of peripheral plasma in Tore Supra tokamak' was carried out as a three-year plan receiving the support of the scientific research expense of the Ministry of Education. This is to apply the method of measuring electron density distribution by neutral lithium beam probe spectroscopy to the measurement of the electron density distribution in the peripheral plasma in Tore Supra Tokamak in France. Among many tokamaks in operation doing respective characteristics researches, the Tore Supra generates the toroidal magnetic field by using superconducting coils, and aims at the long time discharge for 30 sec. for the time being, and for 300 sec. in future. In the plasma generators for long time discharge like this, the technology of particle control is a large problem. For this purpose, a divertor was added to the Tore Supra. In order to advance the research on particle control, it is necessary to examine the behavior of plasma in the peripheral part in detail. The measurement of peripheral plasma in tokamaks, beam probe spectroscopy, the Tore Supra tokamak, the progress of the joint research, the problems in the joint research and the perspective of hereafter are reported. (K.I.)
Electron thermal transport in tokamak plasmas
Energy Technology Data Exchange (ETDEWEB)
Konings, J A
1994-11-30
The process of fusion of small nuclei thereby releasing energy, as it occurs continuously in the sun, is essential for the existence of mankind. The same process applied in a controlled way on earth would provide a clean and an abundant energy source, and be the long term solution of the energy problem. Nuclear fusion requires an extremely hot (10{sup 8} K) ionized gas, a plasma, that can only be maintained if it is kept insulated from any material wall. In the so called `tokamak` this is achieved by using magnetic fields. The termal insulation, which is essential if one wants to keep the plasma at the high `fusion` temperature, can be predicted using basic plasma therory. A comparison with experiments in tokamaks, however, showed that the electron enery losses are ten to hundred times larger than this theory predicts. This `anomalous transport` of thermal energy implies that, to reach the condition for nuclear fusion, a fusion reactor must have very large dimensions. This may put the economic feasibility of fusion power in jeopardy. Therefore, in a worldwide collaboration, physicists study tokamak plasmas in an attempt to understand and control the energy losses. From a scientific point of view, the mechanisms driving anomalous transport are one of the challenges in fudamental plasma physics. In Nieuwegein, a tokamak experiment (the Rijnhuizen Tokamak Project, RTP) is dedicated to the study of anomalous transport, in an international collaboration with other laboratories. (orig./WL).
Neoclassical MHD equations for tokamaks
International Nuclear Information System (INIS)
Callen, J.D.; Shaing, K.C.
1986-03-01
The moment equation approach to neoclassical-type processes is used to derive the flows, currents and resistive MHD-like equations for studying equilibria and instabilities in axisymmetric tokamak plasmas operating in the banana-plateau collisionality regime (ν* approx. 1). The resultant ''neoclassical MHD'' equations differ from the usual reduced equations of resistive MHD primarily by the addition of the important viscous relaxation effects within a magnetic flux surface. The primary effects of the parallel (poloidal) viscous relaxation are: (1) Rapid (approx. ν/sub i/) damping of the poloidal ion flow so the residual flow is only toroidal; (2) addition of the bootstrap current contribution to Ohm's laws; and (3) an enhanced (by B 2 /B/sub theta/ 2 ) polarization drift type term and consequent enhancement of the perpendicular dielectric constant due to parallel flow inertia, which causes the equations to depend only on the poloidal magnetic field B/sub theta/. Gyroviscosity (or diamagnetic vfiscosity) effects are included to properly treat the diamagnetic flow effects. The nonlinear form of the neoclassical MHD equations is derived and shown to satisfy an energy conservation equation with dissipation arising from Joule and poloidal viscous heating, and transport due to classical and neoclassical diffusion
Alfven Eigenmodes in spherical tokamaks
International Nuclear Information System (INIS)
Gryaznevich, Mikhail P.; Sharapov, Sergei E.; Berk, Herbert L.; Pinches, Simon D.
2005-01-01
Electromagnetic instabilities are often excited by fast super-Alfvenic ions produced by neutral beam injection (NBI) in plasmas of the spherical tokamaks START and MAST (toroidal magnetic confinement devices in which the minor a and major R 0 radii of the torus are comparable, R 0 /a≅1.2/1.8). These instabilities are seen as discrete weakly-damped toroidal and elliptical Alfven Eigenmodes (TAEs and EAEs) with frequencies tracing in time the Alfven scaling with the equilibrium magnetic field and plasma density, or as energetic particle modes (EPMs) whose frequencies don't start from TAE-frequency and sweep down in time faster than the equilibrium parameters change. In some discharges the beam drives Aflvenic-type modes that start from the TAE frequency and sweep in both up- and down- directions. Such electromagnetic perturbations are interpreted as 'hole-clump' long-living nonlinear fluctuations of the fast ion distribution function predicted by Berk-Breizman-Petviashvili [Phys. Lett. A238 (1998) 408]. It is found on both START and MAST that the Alfven instabilities weaken in their mode amplitude and in the number of unstable modes as the pressure of the thermal plasma increases, in agreement with increased thermal ion Landau damping and the pressure effect on core-localised TAEs. (author)
The collaborative tokamak control room
International Nuclear Information System (INIS)
Schissel, D.P.
2006-01-01
Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaborations between experimental sites and laboratories worldwide. In the US, the National Fusion Collaboratory Project is developing a persistent infrastructure to enable scientific collaboration for all aspects of magnetic fusion energy research by creating a robust, user-friendly collaborative environment and deploying this to the more than 1000 US fusion scientists in 40 institutions who perform magnetic fusion research. This paper reports on one aspect of the project which is the development of the collaborative tokamak control room to enhance both collocated and remote scientific participation in experimental operations. This work includes secured computational services that can be scheduled as required, the ability to rapidly compare experimental data with simulation results, a means to easily share individual results with the group by moving application windows to a shared display, and the ability for remote scientists to be fully engaged in experimental operations through shared audio, video, and applications. The project is funded by the USDOE Office of Science, Scientific Discovery through Advanced Computing (SciDAC) Program and unites fusion and computer science researchers to directly address these challenges
Tokamak rotation and charge exchange
International Nuclear Information System (INIS)
Hazeltine, R.D.; Rowan, W.L.; Solano, E.R.; Valanju, P.M.
1991-01-01
In the absence of momentum input, tokamak toroidal rotation rates are typically small - no larger in particular than poloidal rotation - even when the radial electric field is strong, as near the plasma edge. This circumstance, contradicting conventional neoclassical theory, is commonly attributed to the rotation damping effect of charge exchange, although a detailed comparison between charge-exchange damping theory and experiment is apparently unavailable. Such a comparison is attempted here in the context of recent TEXT experiments, which compare rotation rates, both poloidal and toroidal, in helium and hydrogen discharges. The helium discharges provide useful data because they are nearly free of ion-neutral charge exchange; they have been found to rotate toroidally in reasonable agreement with neoclassical predictions. The hydrogen experiments show much smaller toroidal motion as usual. The theoretical calculation uses the full charge-exchange operator and assumes plateau collisionality, roughly consistent with the experimental conditions. The authors calculate the ion flow as a function of v cx /v c , where v cx is the charge exchange rate and v c the Coulomb collision frequency. The results are in reasonable accord with the observations. 1 ref
International Nuclear Information System (INIS)
Isayama, Akihiko; Tanaka, Hitoshi
2009-01-01
Recent progress in the ECH·ECCD study in tokamak and spherical tokamak devices is described. As for the tokamak study, results on the control of neoclassical tearing modes and sawtooth oscillations, the current profile, the internal transport barrier, the plasma start-up and the discharge cleaning are given. As for the spherical tokamak study, the plasma start-up by ECH·ECCD and the electron-Bernstein-wave heating and the current drive are described. (T.I.)
Beta limit of crescent and bean shaped tokamaks
International Nuclear Information System (INIS)
Naitou, H.; Yamazaki, K.
1988-01-01
The maximum attainable beta values which can be expected in tokamaks with crescent (BEAN 1) and rounded (BEAN 2) bean shaped cross-sections are obtained numerically by using the linear ideal MHD stability analysis code ERATO. The current profiles are optimized with a fixed pressure profile for high values of beta, keeping Mercier, high-n ballooning and n=1 kink modes stable. The poloidal plasma cross-sections are inscribed in a rectangle with an aspect ratio of three and an ellipticity of two. A confocal wall, the distance of which from the plasma surface is equal to the horizontal minor plasma radius, is present to stabilize against the kink mode. Depending on the shape and triangularity (indentation), a beta value of 10 to 17% is obtained. It is also shown that the coefficient of the Troyon-type beta scaling increases for an indented plasma. In the case of small indentation, the BEAN 1 type tokamaks show higher beta values than the BEAN 2 type. For strong indentation, the BEAN 2 type gives the highest beta value. (author). 29 refs, 15 figs
Kinetic theory of plasma adiabatic major radius compression in tokamaks
International Nuclear Information System (INIS)
Gorelenkova, M.V.; Gorelenkov, N.N.; Azizov, E.A.; Romannikov, A.N.; Herrmann, H.W.
1998-01-01
In order to understand the individual charged particle behavior as well as plasma macroparameters (temperature, density, etc.) during the adiabatic major radius compression (R-compression) in a tokamak, a kinetic approach is used. The perpendicular electric field from the Ohm close-quote s law at zero resistivity is made use of in order to describe particle motion during the R-compression. Expressions for both passing and trapped particle energy and pitch angle change are derived for a plasma with high aspect ratio and circular magnetic surfaces. The particle behavior near the passing trapped boundary during the compression is studied to simulate the compression-induced collisional losses of alpha particles. Qualitative agreement is obtained with the alphas loss measurements in deuterium-tritium (D-T) experiments in the Tokamak Fusion Test Reactor (TFTR) [World Survey of Activities in Controlled Fusion Research [Nucl. Fusion special supplement (1991)] (International Atomic Energy Agency, Vienna, 1991)]. The plasma macroparameters evolution at the R-compression is calculated by solving the gyroaveraged drift kinetic equation. copyright 1998 American Institute of Physics
Feasibility study of a fission-suppressed tokamak fusion breeder
International Nuclear Information System (INIS)
Moir, R.W.; Lee, J.D.; Neef, W.S.
1984-12-01
The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m 2 and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of 233 U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U 3 O 8 depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management
Concept study of the Steady State Tokamak Reactor (SSTR)
International Nuclear Information System (INIS)
1991-06-01
The Steady State Tokamak Reactor (SSTR) concept has been proposed as a realistic fusion power reactor to be built in the near future. An overall concept of SSTR is introduced which is based on a small extension of the present day physics and technologies. The major feature of SSTR is the maximum utilization of a bootstrap current in order to reduce the power required for the steady state operation. This requirement leads to the choice of moderate current (12 MA), and high βp (2.0) for the device, which are achieved by selecting high aspect ratio (A=4) and high toroidal magnetic field (16.5 T). A negative-ion-based neutral beam injection system is used both for heating and central current drive. Notable engineering features of SSTR are: the use of a uniform vacuum vessel and periodical replacements of the first wall and blanket layers and significant reduction of the electromagnetic force with the use of functionally gradient material. It is shown that a tokamak machine comparable to ITER in size can become a power reactor capable of generating about 1 GW of electricity with a plant efficiency of ∼30%. (author)
Electromagnetic effects on trace impurity transport in tokamak plasmas
Hein, T.; Angioni, C.
2010-01-01
The impact of electromagnetic effects on the transport of light and heavy impurities in tokamak plasmas is investigated by means of an extensive set of linear gyrokinetic numerical calculations with the code GYRO [J. Candy and R. E. Waltz, J. Comput. Phys. 186, 545 (2003)] and of analytical derivations with a fluid model. The impurity transport is studied by appropriately separating diffusive and convective contributions, and conditions of background microturbulence dominated by both ion temperature gradient (ITG) and trapped electron modes (TEMs) are analyzed. The dominant contribution from magnetic flutter transport turns out to be of pure convective type. However it remains small, below 10% with respect to the E ×B transport. A significant impact on the impurity transport due to an increase in the plasma normalized pressure parameter β is observed in the case of ITG modes, while for TEM the overall effect remains weak. In realistic conditions of high β plasmas in the high confinement (H-) mode with dominant ITG turbulence, the impurity diffusivity is found to decrease with increasing β in qualitative agreement with recent observations in tokamaks. In contrast, in these conditions, the ratio of the total off-diagonal convective velocity to the diagonal diffusivity is not strongly affected by an increase in β, particularly at low impurity charge, due to a compensation between the different off-diagonal contributions.
Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code
International Nuclear Information System (INIS)
Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P
2004-01-01
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values
Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code
Energy Technology Data Exchange (ETDEWEB)
Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)
2004-09-01
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.
Heat load material studies: Simulated tokamak disruptions
International Nuclear Information System (INIS)
Gahl, J.M.; McDonald, J.M.; Zakharov, A.; Tserevitinov, S.; Barabash, V.; Guseva, M.
1991-01-01
It is clear that an improved understanding of the effects of tokamak disruptions on plasma facing component materials is needed for the ITER program. very large energy fluxes are predicted to be deposited in ITER and could be very damaging to the machine. During 1991, Sandia National Laboratories and the University of New Mexico conducted cooperative tokamak disruption simulation experiments at several Soviet facilities. These facilities were located at the Efremov Institute in Leningrad, the Kurchatov Atomic Energy Institute (Troisk and Moscow) and the Institute for Physical Chemistry of the Soviet Adademy of Sciences in Moscow. Erosion of graphite from plasma stream impact is seen to be much less than that observed with laser or electron beams with similar energy fluxes. This, along with other data obtained, seem to suggest that the ''vapor shielding'' effect is a very important phenomenon in the study of graphite erosion during tokamak disruption
Effect of impurity radiation on tokamak equilibrium
International Nuclear Information System (INIS)
Rebut, P.H.; Green, B.J.
1977-01-01
The energy loss from a tokamak plasma due to the radiation from impurities is of great importance in the overall energy balance. Taking the temperature dependence of this loss for two impurities characteristic of those present in existing tokamak plasmas, the condition for radial power balance is derived. For the impurities considered (oxygen and iron) it is found that the radiation losses are concentrated in a thin outer layer of the plasma and the equilibrium condition places an upper limit on the plasma paraticle number density in this region. This limiting density scales with mean current density in the same manner as is experimentally observed for the peak number density of tokamak plasmas. The stability of such equilibria is also discussed. (author)
Spherical tokamak power plant design issues
International Nuclear Information System (INIS)
Hender, T.C.; Bond, A.; Edwards, J.; Karditsas, P.J.; McClements, K.G.; Mustoe, J.; Sherwood, D.V.; Voss, G.M.; Wilson, H.R.
2000-01-01
The very high β potential of the spherical tokamak has been demonstrated in the START experiment. Systems code studies show the cost of electricity from spherical tokamak power plants, operating at high β in second ballooning mode stable regime, is comparable with fossil fuels and fission. Outline engineering designs are presented based on two concepts for the central rod of the toroidal field (TF) circuit - a room temperature water cooled copper rod or a helium cooled cryogenic aluminium rod. For the copper rod case the TF return limbs are supported by the vacuum vessel, while for the aluminium rod the TF coils form an independent structure. In both cases thermohydraulic and stress calculations indicate the viability of the design. Two-dimensional neutronics calculations show the feasibility of tritium self-sufficiency without an inboard blanket. The spherical tokamak has unique maintenance possibilities based on lowering major component structures into a hot cell beneath the device and these are discussed
Time - resolved thermography at Tokamak T-10
International Nuclear Information System (INIS)
Grunow, C.; Guenther, K.; Lingertat, J.; Chicherov, V.M.; Evstigneev, S.A.; Zvonkov, S.N.
1987-01-01
Thermographic experiments were performed at T-10 tokamak to investigate the thermic coupling of plasma and the limiter. The limiter is an internal equipment of the vacuum vessel of tokamak-type fusion devices and the interaction of plasma with limiter results a high thermal load of limiter for short time. In according to improve the limiter design the temperature distribution on the limiter surface was measured by a time-resolved thermographic method. Typical isotherms and temperature increment curves are presented. This measurement can be used as a systematic plasma diagnostic method because the limiter is installed in the tokamak whereas special additional probes often disturb the plasma discharge. (D.Gy.) 3 refs.; 7 figs
Magnet design considerations for Tokamak fusion reactors
International Nuclear Information System (INIS)
Purcell, J.R.; Chen, W.; Thomas, R.
1976-01-01
Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)
Ripple induced trapped particle loss in tokamaks
International Nuclear Information System (INIS)
White, R.B.
1996-05-01
The threshold for stochastic transport of high energy trapped particles in a tokamak due to toroidal field ripple is calculated by explicit construction of primary resonances, and a numerical examination of the route to chaos. Critical field ripple amplitude is determined for loss. The expression is given in magnetic coordinates and makes no assumptions regarding shape or up-down symmetry. An algorithm is developed including the effects of prompt axisymmetric orbit loss, ripple trapping, convective banana flow, and stochastic ripple loss, which gives accurate ripple loss predictions for representative Tokamak Fusion Test Reactor and International Thermonuclear Experimental Reactor equilibria. The algorithm is extended to include the effects of collisions and drag, allowing rapid estimation of alpha particle loss in tokamaks
Activation analysis of the compact ignition tokamak
International Nuclear Information System (INIS)
Selcow, E.C.
1986-01-01
The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak
Helicity content and tokamak applications of helicity
International Nuclear Information System (INIS)
Boozer, A.H.
1986-05-01
Magnetic helicity is approximately conserved by the turbulence associated with resistive instabilities of plasmas. To generalize the application of the concept of helicity, the helicity content of an arbitrary bounded region of space will be defined. The definition has the virtues that both the helicity content and its time derivative have simple expressions in terms of the poloidal and toroidal magnetic fluxes, the average toroidal loop voltage and the electric potential on the bounding surface, and the volume integral of E-B. The application of the helicity concept to tokamak plasmas is illustrated by a discussion of so-called MHD current drive, an example of a stable tokamak q profile with q less than one in the center, and a discussion of the possibility of a natural steady-state tokamak due to the bootstrap current coupling to tearing instabilities
Heavy ion beam probe development for the plasma potential measurement on the TUMAN-3M tokamak
International Nuclear Information System (INIS)
Askinazi, L.G.; Kornev, V.A.; Lebedev, S.V.; Tukachinsky, A.S.; Zhubr, N.A.; Dreval, N.B.; Krupnik, L.I.
2004-01-01
The peculiarities of the heavy ion beam probe implementation on the small aspect ratio tokamak TUMAN-3M are analyzed. The toroidal displacement of beam trajectory due to the high I pl /B tor ratio is taken into account when designing the layout of the diagnostic. Numerical calculation of beam trajectories using realistic configuration of TUMAN-3M magnetic fields and parabolic plasma current profile resulted in proper adjustment of probing and detection parameters (probing ion material, energy, entrance angles, detector location, and orientation). Secondary ion energy analyzer gain functions G and F were measured in situ using neutral hydrogen puffed in the tokamak vessel as a target for secondary ions production. The detector unit featured split-plate design and had additional electrodes for secondary electron emission suppression. As a result, the diagnostic is now capable of plasma potential evolution measurement and is sensitive enough to trace the potential profile evolution at the L-H mode transition
Proposed tokamak poloidal field system development program
Energy Technology Data Exchange (ETDEWEB)
Rogers, J.D.; Vogel, H.F.; Warren, R.W.; Weldon, D.M.
1977-05-01
A program is proposed to develop poloidal field components for TNS and EPR size tokamak devices and to test these components in realistic circuits. Emphasis is placed upon the development of the most difficult component, the superconducting ohmic-heating coil. Switches must also be developed for testing the coils, and this switching technology is to be extended to meet the requirements for the large scale tokamaks. Test facilities are discussed; power supplies, including a homopolar to drive the coils, are considered; and poloidal field systems studies are proposed.
Tokamak Engineering Technology Facility scoping study
Energy Technology Data Exchange (ETDEWEB)
Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.
1976-03-01
A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR.
Ballooning stable high beta tokamak equilibria
International Nuclear Information System (INIS)
Tuda, Takashi; Azumi, Masafumi; Kurita, Gen-ichi; Takizuka, Tomonori; Takeda, Tatsuoki
1981-04-01
The second stable regime of ballooning modes is numerically studied by using the two-dimensional tokamak transport code with the ballooning stability code. Using the simple FCT heating scheme, we find that the plasma can locally enter this second stable regime. And we obtained equilibria with fairly high beta (β -- 23%) stable against ballooning modes in a whole plasma region, by taking into account of finite thermal diffusion due to unstable ballooning modes. These results show that a tokamak fusion reactor can operate in a high beta state, which is economically favourable. (author)
Energy storage for tokamak reactor cycles
International Nuclear Information System (INIS)
Buchanan, C.H.
1979-01-01
The inherent characteristic of a tokamak reactor requiring periodic plasma quench and reignition introduces the problem of energy storage to permit continuous electrical output to the power grid. The cycle under consideration in this paper is a 1000 second burn followed by a 100 second reignition phase. The physical size of a typical toroidal plasma reaction chamber for a tokamak reactor has been described earlier. The thermal energy storage requirements described in this reference will serve as a basis for much of the ensuing discussion
Tokamak power systems studies, FY 1985
International Nuclear Information System (INIS)
Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.
1985-12-01
The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs
A Fast Shutdown Technique for Large Tokamaks
International Nuclear Information System (INIS)
Fredrickson, E.; Schmidt, G.L.; Hill, K.; Jardin, S.C.
1999-01-01
A practical method is proposed for the fast shutdown of a large ignited tokamak. The method consists of injecting a rapid series of 30-50 deuterium pellets doped with a small ( 0.0005%) concentration of Krypton impurity, and simultaneously ramping the plasma current and shaping fields down over a period of several seconds using the poloidal field system. Detailed modeling with the Tokamak Simulation Code using a newly developed pellet mass deposition model shows that this method should terminate the discharge in a controlled and stable way without producing significant numbers of runaway electrons. A partial prototyping of this technique was accomplished in TFTR
Two-ion ICRF heating in Tokamaks
International Nuclear Information System (INIS)
Tennfors, E.
1985-03-01
The practical consequences for tokamak plasma heating in the ion cyclotron frequency regime of the two-dimensional treatment of the two-ion mode conversion layer are analyzed. The problem of evaluation of the condition for fast wave resonance is analyzed, as well as the limitations imposed by warm plasma effects. Simple ways to find the mode conversion surfaces when they exist are presented. Also for large tokamaks, it is possible to obtain mode conversion conditions for realistic antenna spectra provided species concentration and frequency are chosen such that the surface Epsilon = 0 intersects the plasma midplane just outside of the magnetic axis. (Author)
Radial electric fields for improved tokamak performance
International Nuclear Information System (INIS)
Downum, W.B.
1981-01-01
The influence of externally-imposed radial electric fields on the fusion energy output, energy multiplication, and alpha-particle ash build-up in a TFTR-sized, fusing tokamak plasma is explored. In an idealized tokamak plasma, an externally-imposed radial electric field leads to plasma rotation, but no charge current flows across the magnetic fields. However, a realistically-low neutral density profile generates a non-zero cross-field conductivity and the species dependence of this conductivity allows the electric field to selectively alter radial particle transport
Tokamak power systems studies, FY 1985
Energy Technology Data Exchange (ETDEWEB)
Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Smith, D.L.; Sze, D.K.
1985-12-01
The Tokamak Power System Studies (TPSS) at ANL in FY-1985 were devoted to exploring innovative design concepts which have the potential for making substantial improvements in the tokamak as a commercial power reactor. Major objectives of this work included improved reactor economics, improved environmental and safety features, and the exploration of a wide range of reactor plant outputs with emphasis on reduced plant sizes compared to STARFIRE. The activities concentrated on three areas: plasma engineering, impurity control, and blanket/first wall/shield technology. 205 refs., 125 figs., 107 tabs.
Thermonuclear ignition in the next generation tokamaks
International Nuclear Information System (INIS)
Johner, J.
1989-04-01
The extrapolation of experimental rules describing energy confinement and magnetohydrodynamic - stability limits, in known tokamaks, allow to show that stable thermonuclear ignition equilibria should exist in this configuration, if the product aB t x of the dimensions by a magnetic-field power is large enough. Quantitative application of this result to several next-generation tokamak projects show that those kinds of equilibria could exist in such devices, which would also have enough additional heating power to promote an effective accessible ignition
Shielding and maintainability in an experimental tokamak
International Nuclear Information System (INIS)
Abdou, M.A.; Fuller, G.; Hager, E.R.; Vogelsang, W.F.
1979-01-01
This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue
Periodic disruptions in the MT-1 tokamak
International Nuclear Information System (INIS)
Zoletnik, S.
1988-11-01
Disruptive instabilities are common phenomena in toroidal devices, especially in tokamaks. Three types can be distinguished: internal, minor and major disruptions. Periodic minor disruptions in the MT-1 tokamak were measured systematically with values of the limiter safety factor between 4 and 10. The density limit as a function of plasma current and horizontal displacement was investigated. Precursor oscillations always appear before the instability with increasing amplitude but can be observed at the density limit with quasi-stationary amplitude. Phase correlation between precursor oscillations were measured with Mirnov coils and x-ray detectors, and they show good agreement with a simple magnetic island model. (R.P.) 11 refs.; 6 figs
Gas blanket fueling of a tokamak reactor
International Nuclear Information System (INIS)
Gralnick, S.L.
1978-01-01
The purpose of this paper is a speculative investigation of the potential of fueling a Tokamak by introducing a sufficiently large quantity of gaseous deuterium and tritium at the vacuum wall boundary. It is motivated by two factors: current generation tokamaks are, in a manner of speaking, fueled from the edge quite successfully as is evidenced by pulse lengths that are long compared to particle recycling times, and by rapid plasma density increase produced by gas puffing, alternative, deep penetration fueling techniques that have been proposed possess severe technological problems and large costs
Electronic system of TBR tokamak device
International Nuclear Information System (INIS)
Silva, R.P. da.
1980-01-01
The electronics developed as a part of the TBR project, which involves the construction of a small tokamak at the Physics Institute of the University of Sao Paulo, is described. On the basis of tokamak parameter values, the electronics for the toroidal field, ohmic/heating and vertical field systems is presented, including capacitors bank, switches, triggering circuits and power supplies. A controlled power oscilator used in discharge cleaning and pre-ionization is also described. The performance of the system as a function of the desired plasma parameters is discussed. (Author) [pt
Tokamak Engineering Technology Facility scoping study
International Nuclear Information System (INIS)
Stacey, W.M. Jr.; Abdou, M.A.; Bolta, C.C.
1976-03-01
A scoping study for a Tokamak Engineering Technology Facility (TETF) is presented. The TETF is a tokamak with R = 3 m and I/sub p/ = 1.4 MA based on the counterstreaming-ion torus mode of operation. The primary purpose of TETF is to demonstrate fusion technologies for the Experimental Power Reactor (EPR), but it will also serve as an engineering and radiation test facility. TETF has several technological systems (e.g., superconducting toroidal-field coil, tritium fuel cycle, impurity control, first wall) that are prototypical of EPR
Alfven wave heating in a tokamak reactor
International Nuclear Information System (INIS)
Borg, G.G.; Appert, K.; Knight, A.J.; Lister, J.B.; Vaclavik, J.
1990-01-01
A number of features of Alfven wave heating make it potentially attractive for use in large tokamak reactors. Among them are the availability and relativity low cost of the power supplies, the potential ability to act selectively on the current profile, and the probable absence of operational limits in size, fields or density. The physics of Alfven wave heating in a large tokamak is assessed. Present theoretical understanding of mode coupling and antenna loading is extrapolated to a large machine. The problem of a recessed antenna is analysed. Calculations of loading and discussion of various heating scenarios for the particular case of NET are also presented. (author). 23 refs, 18 figs, 4 tabs
Can better modelling improve tokamak control?
International Nuclear Information System (INIS)
Lister, J.B.; Vyas, P.; Ward, D.J.; Albanese, R.; Ambrosino, G.; Ariola, M.; Villone, F.; Coutlis, A.; Limebeer, D.J.N.; Wainwright, J.P.
1997-01-01
The control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed. (author) 5 figs., 9 refs
Increase in beta limit in tokamak plasmas
International Nuclear Information System (INIS)
Kamada, Yutaka
2003-01-01
This paper reviews recent studies of tokamak MHD stability towards the achievement of a high beta steady-state, where the profile control of current, pressure, and rotation, and the optimization of the plasma shape play fundamental roles. The key instabilities include the neoclassical tearing mode, the resistive wall mode, the edge localized mode, etc. In order to demonstrate an economically attractive tokamak reactor, it is necessary to increase the beta value simultaneously with a sufficiently high integrated plasma performance. Towards this goal, studies of stability control in self-regulating plasma systems are essential. (author)
Transport in a small aspect ratio torus
International Nuclear Information System (INIS)
White, R.B.; Gates, D.A.; Mynick, H.E.
2005-01-01
Transport theory in toroidal devices often assumes large aspect ratio and also assumes the poloidal field is small compared to the toroidal field. These assumptions result in transport which in the low collision rate limit is dominated by banana orbits, giving the largest collisionless excursion of a particle from an initial flux surface. However in a small aspect ratio device the gyro radius may be larger than the banana excursion, resulting in significant deviations from the standard neoclassical predictions. In this paper we report numerical simulation of diffusion in low and high beta low aspect ratio equilibria. We also sketch an analytic theory. The diffusion, which we refer to as omniclassical, is a combination of neoclassical and properly averaged classical effects, and can be two or three times the neoclassical value. Good agreement of the analytic theory with numerical simulations is obtained. (author)
Stability of n = 1 kink modes in bean-shaped tokamaks
International Nuclear Information System (INIS)
Manickam, J.; Grimm, R.C.; Okabayashi, M.
1983-08-01
Numerical studies show that by indenting the small-major-radius side of conventional finite-aspect-ratio tokamaks, significant improvements to the stability of pressure-driven ideal MHD modes can be achieved. The internal n - 1 kink mode can be stabilized completely with quite modest indentation. Kink-ballooning mode stability is also improved, and, in the presence of a nearby conducting wall, accessibility to a second stable region at high plasma β is possible
Energy Technology Data Exchange (ETDEWEB)
Liu, X.; Zhao, H. L.; Liu, Y., E-mail: liuyong@ipp.ac.cn; Li, E. Z.; Han, X.; Ti, A.; Hu, L. Q.; Zhang, X. D. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Domier, C. W.; Luhmann, N. C. [Department of Electrical and Computer Engineering, University of California at Davis, Davis, California 95616 (United States)
2014-09-15
This paper presents the results of the in situ absolute intensity calibration for the 32-channel heterodyne radiometer on the experimental advanced superconducting tokamak. The hot/cold load method is adopted, and the coherent averaging technique is employed to improve the signal to noise ratio. Measured spectra and electron temperature profiles are compared with those from an independent calibrated Michelson interferometer, and there is a relatively good agreement between the results from the two different systems.
Maximum attainable power density and wall load in tokamaks underlying reactor relevant constraints
International Nuclear Information System (INIS)
Borrass, K.; Buende, R.
1979-09-01
The characteristic data of tokamaks optimized with respect to their power density or wall load are determined. Reactor relevant constraints are imposed, such as a fixed plant net power output, a fixed blanket thickness and the dependence of the maximum toroidal field on the geometry and conductor material. The impact of finite burn times is considered. Various scaling laws of the toroidal beta with the aspect ratio are discussed. (orig.) 891 GG/orig. 892 RDG [de
International Nuclear Information System (INIS)
Sheffield, J.
1985-01-01
During recent years improvements have been made to tokamak startup procedures, which are important to the optimization of ignited tokamaks. The use of rf-assisted startup and noninductive current drive has led to substantial reduction and even complete elimination of the volt-seconds used during startup, relaxing constraints on poloidal coil, vacuum vessel, and structure design. This paper reviews these and other improvements and discusses the various bulk heating techniques that may be used to ignite a D-T plasma
A need for non-tokamak approaches to magnetic fusion energy
International Nuclear Information System (INIS)
Bathke, C.G.; Krakowski, R.A.; Miller, R.L.
1992-01-01
Focusing exclusively on conventional tokamak physics in the quest for commercial fusion power is premature, and the options for both advanced-tokamak and non-tokamak concepts need continued investigation. The basis for this claim is developed, and promising advanced-tokamak and non-tokamak options are suggested
Ballooning mode second stability region for sequences of tokamak equilibria
International Nuclear Information System (INIS)
Sugiyama, L.; Mark, J.W.K.
A numerical study of several sequences of tokamak equilibria derived from two flux conserving sequences confirms the tendency of high n ideal MHD ballooning modes to stabilize for values of the plasma beta greater than a second critical beta, for sufficiently favorable equilibria. The major stabilizing effect of increasing the inverse rotational transform profile q(Psi) for equilibria with the same flux surface geometry is shown. The unstable region shifts toward larger shear d ln q/d ln γ and the width of the region measured in terms of the poloidal beta or a pressure gradient parameter, for fixed shear, decreases. The smaller aspect ratio sequences are more sensitive to changes in q and have less stringent limits on the attainable value of the plasma beta in the high beta stable region. Finally, the disconnected mode approximation is shown to provide a reasonable description of the second high beta stability boundary
A consistency analysis on the tokamak reactor plasmas
International Nuclear Information System (INIS)
Fukuyama, A.; Itoh, S.-I.; Itoh, K.
1990-12-01
The parameter regime which simultaneously fulfills the various physics constraints are looked for in the case of ITER grade tokamaks. The consistency analysis code is applied. It is found that, if the energy confinement time reaches 1.6 times of the prediction of the L-mode scaling law, the Q-value of about 4 is possible for the full current drive operation at the input power P in of 100MW (Q is the ratio of fusion output and P in ). In the ignition mode, where half of the current is inductively sustained, Q approaches to 15 for this circulating power. If only the L-mode is realized, Q is about 1.5 for P in ≅100 MW. (author)
Tokamak fusion reactors with less than full tritium breeding
International Nuclear Information System (INIS)
Evans, K. Jr.; Gilligan, J.G.; Jung, J.
1983-05-01
A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed
Tokamak power reactor ignition and time dependent fractional power operation
International Nuclear Information System (INIS)
Vold, E.L.; Mau, T.K.; Conn, R.W.
1986-06-01
A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve
MHD stability properties of bean-shaped tokamaks
International Nuclear Information System (INIS)
Grimm, R.C.; Chance, M.S.; Todd, A.M.M.
1984-03-01
A study of the MHD stability properties of bean-shaped tokamak plasmas is presented. For ballooning modes, while increased indentation gives larger β stable configurations, the existence and accessibility of the second stable region is sensitive to the pressure and safety factor profiles. The second stable region appears at lower β values for large aspect ratio and moderately high q-values. Finite-Larmor-radius (FLR) kinetic effects can significantly improve the stability properties. For low q (< 1) operation, long wavelength (n approx. 2,3) internal pressure driven modes occur at modest β/sub p/ values and accessibility to higher β operation is unlikely. Indentation modifies the nature of the usually vertical axisymmetric instability, but the mode can be passively stabilized by placing highly conducting plates near to the tips of the plasma bean. At constant q, indentation has a stabilizing effect on tearing modes
Drift-kinetic Alfven modes in high performance tokamaks
International Nuclear Information System (INIS)
Jaun, A.; Fasoli, A.F.; Testa, D.; Vaclavik, J.; Villard, L.
2001-01-01
The stability of fast-particle driven Alfven eigenmodes is modeled in high performance tokamaks, successively with a conventional shear, an optimized shear and a tight aspect ratio plasma. A large bulk pressure yields global kinetic Alfven eigenmodes that are stabilized by mode conversion in the presence of a divertor. This suggests how conventional reactor scenarii could withstand significant pressure gradients from the fusion products. A large safety factor in the core q 0 >2.5 in deeply shear reversed configurations and a relatively large bulk ion Larmor radius in a low magnetic field can trigger global drift-kinetic Alfven eigenmodes that are unstable in high performance JET, NSTX and ITER plasmas. (author)
Positional instability analysis of tokamak plasmas by ERATO
International Nuclear Information System (INIS)
Kumagai, Michikazu; Tsunematsu, Toshihide; Tokuda, Shinji; Takeda, Tatsuoki
1983-06-01
The stability of axisymmetric modes of a tokamak plasma(positional instabilities) is analyzed for the Solov'ev equilibrium by using the linear ideal MHD code ERATO-J. The dependence of the stability on various parameters, i.e., the ellipticity and triangularity of the plasma cross-section, the aspect ratio, the safety factor at the magnetic axis, and the distance between the plasma and a conducting shell is investigated. Comparison of the results with those by the rigid model shows that the stability condition derived from the rigid model in terms of the decay index(n-index) of the external equilibrating field is a good approximation for the plasma with small triangular deformation. Also the results are compared with those of the rigid displacement model and applicability of the various models on the positional instability analyses is discussed. (author)
Compact Ignition Tokamak Program: R and D needs
International Nuclear Information System (INIS)
Flanagan, C.A.
1985-01-01
This report on the Compact Ignition Tokamak Program supplies information concerning: segmented vacuum vessel joint development; first wall tile attachments; first wall/tile development - composite materials; vacuum leak detection; high frequency rf sources; Faraday shield development; design and testing of rf launchers for high power, ling pulse operation; radiation hardened, low loss, dielectric windows for rf, IR, visible, UV and X-rays, mirrors for changing direction and focusing IR, visible and UV radiation; radiation resistant optical dielectric wave guides; radiation resistant HV insulation for diagnostic magnetic pickup coils; compact radiation and/or magnetic shielding for in-vault diagnostics that need some attenuation to reduce S/N ratio; radiation hardened line-of-sight sensors such as bolometers, UV and soft X-ray detectors, neutral particle analyzers, torus pressure gauges; special maintenance fixtures and tools; material properties - design data base - all materials; and insulation - electrical/thermal and mechanical properties
Overview and initial results of the ETE spherical tokamak
International Nuclear Information System (INIS)
Berni, L.A.; Del Bosco, E.; Ferreira, J.G.; Ludwig, G.O.; Oliveira, R.M.; Shibata, C.S.; Barbosa, L.F.F.P.W.; Vilela, W.A.
2003-01-01
The ETE spherical tokamak is a small size aspect-ratio machine with major and minor radius of 30 cm and 20 cm, respectively. The vessel was made of Inconel 625 and provides good access for plasma diagnostics through 58 Conflat ports. The first plasma was obtained at the end of 2000 and presently plasma currents of about 45 kA lasting for about 4 ms with electron temperature up to 160 eV and densities of 2.2x10 19 m -3 are routinely obtained. Achievement of the designed parameters for the first phase of operation is expected by the end of this year, with plasma current up to 200 kA lasting for about 15 ms. This paper describes some details of the ETE project, construction and mainly the first results and analysis of basic parameters. (author)
Internal helical modes with m > 1 in a tokamak with a small shear and high plasma pressure
International Nuclear Information System (INIS)
Mikha lovskij, A.B.; Aburdzhaniya, G.D.; Krymskij, A.M.
1979-01-01
Internal helical modes with m>1 in a circular cross-section tokamak with a small shear and large value of the parameter β (β is the ratio between the mean plasma pressure and the mean pressure of the poloidal magnetic field) are investigated. The equations obtained are used to study the destabilizing effects leading to helical instabilities. The role of destabilizing effects is regarded both in local and in a nonlocal approximations on the assumption that the radial plasma pressure is distributed parabolically and that the radial current distribution is also parabolic though slightly varying. It has been established that the profiling of current may lead to the tokamak plasma stability with respect to the modes under investigation. A tokamak with a small shear has been shown to be more stable relative to these modes than that with a large shear
Tokamak startup with electron cyclotron heating
International Nuclear Information System (INIS)
Holly, D.J.; Prager, S.C.; Shepard, D.A.; Sprott, J.C.
1980-04-01
Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed
Transformer Recharging with Alpha Channeling in Tokamaks
International Nuclear Information System (INIS)
Fisch, N.J.
2009-01-01
Transformer recharging with lower hybrid waves in tokamaks can give low average auxiliary power if the resistivity is kept high enough during the radio frequency (rf) recharging stage. At the same time, operation in the hot ion mode via alpha channeling increases the effective fusion reactivity. This paper will address the extent to which these two large cost saving steps are compatible.
Criteria for initiation of tokamak disruptions
International Nuclear Information System (INIS)
Hopcraft, K.I.; Turner, M.F.
1986-01-01
The process by which a tokamak plasma evolves from an equilibrium state containing a saturated magnetic island to one which is disruptively unstable is discussed and illustrated by numerical simulation of a resistive magnetoplasma. Those elements which are required to initiate a disruption are delineated
Feedback stabilization of axisymmetric modes in tokamaks
International Nuclear Information System (INIS)
Jardin, S.C.; Larrabee, D.A.
1982-01-01
Noncircular tokamak plasmas can be unstable to ideal MHD axisymmetric instabilities. Passive conductors with finite resistivity will at best slow down these instabilities to the resistive (L/R) time of the conductors. An active feedback system far from the plasma which responds on this resistive time can stabilize the system provided its mutual inductance with the passive coils is small enough
Scattering measurements in Tokamak type devices
International Nuclear Information System (INIS)
Matoba, Tohru
1975-03-01
Theories, experiments and proposals for light scattering in Tokamak type devices are reviewed. Thomson scattering, measuring method of the current density distribution by scattering and resonance fluorescence are summarily described. These methods may be useful for diagnosis of the fusion plasmas. The report may help planning of the measuring apparatus for the fusion plasmas in future. (auth.)
Maintenance considerations of the STARFIRE commercial tokamak
International Nuclear Information System (INIS)
Trachsel, C.A.; Zahn, H.S.; Field, R.E.; Stevens, H.C.
1979-01-01
This paper presents the maintenance approach, the commercial tokamak design features that enhance maintenance and preliminary repair time and required mean-time-between-failures for major subsystems. Reactor hall building and maintenance equipment requirements including hot cells, coil rewinding, and cranes are discussed
Tokamak fusion test reactor. Final design report
International Nuclear Information System (INIS)
1978-08-01
Detailed data are given for each of the following areas: (1) system requirements, (2) the tokamak system, (3) electrical power systems, (4) experimental area systems, (5) experimental complex, (6) neutral beam injection system, (7) diagnostic system, and (8) central instrumentation control and data acquisition system
Tokamak power plant burn cycle options
International Nuclear Information System (INIS)
Ehst, D.A.
1994-06-01
Experiments show that tokamaks can operate in various fashions. Economic analyses show that steady state is most attractive provided the physics and technology of current drive (CD) can be modestly improved. Even with very conservative CD assumptions a hybrid operating mode seems superior to conventional, simple inductive operation
Plasma-gun fueling for tokamak reactors
International Nuclear Information System (INIS)
Ehst, D.A.
1980-11-01
In light of the uncertain extrapolation of gas puffing for reactor fueling and certain limitations to pellet injection, the snowplow plasma gun has been studied as a fueling device. Based on current understanding of gun and plasma behavior a design is proposed, and its performance is predicted in a tokamak reactor environment
Preliminary measurements on Tokamak KT-5
International Nuclear Information System (INIS)
Wen Yizhi; Wan Shude; Rong Furui; Haan Shengshen; Liu Wandong; Liu Lei
1987-01-01
A small tokamak, KT-5, has been put in to operation since 1984. The major and minor radius of the plasma are 30 and 4.5 cm, respectively. The parameters obtained in the first phase of KT-5 experiments are as follows B t = 0.45 T, I p ≥ 5 kA, q(α) σ = 50 eV
Experimental methods to study tokamak plasma stability
International Nuclear Information System (INIS)
Perez-Navarro, A.
1978-01-01
Experimental devices to measure external instability modes with small pick-up coils to detect poloidal magnetic field fluctuations, and internal modes with soft-X-ray detectors are discussed. The characteristics of these devices are calculated for a small tokamak (R 0 = 30 cm, a = 10 cm, I 0 50 KA). (author)
MHD stability of vertically asymmetric tokamak equilibria
International Nuclear Information System (INIS)
Dalhed, H.E.; Grimm, R.C.; Johnson, J.L.
1981-03-01
The ideal MHD stability properties of a special class of vertically asymmetric tokamak equilibria are examined. The calculations confirm that no major new physical effects are introduced and the modifications can be understood by conventional arguments. The results indicate that significant departures from up-down symmetry can be tolerated before the reduction in β becomes important for reactor operation
[High beta tokamak research and plasma theory
International Nuclear Information System (INIS)
1990-01-01
Our activities on High Beta Tokamak Research during the past 12 months of the present budget period can be divided into four areas: completion of kink mode studies in HBT; completion of carbon impurity transport studies in HBT; design of HBT-EP; and construction of HBT-EP. Each of these is described briefly in the sections of this progress report
Computation of tokamak equilibria with steady flow
International Nuclear Information System (INIS)
Kerner, W.; Tokuda, Shinji
1987-08-01
The equations for ideal MHD equilibria with stationary flow are reexamined and addressed as numerically applied to tokamak configurations with a free plasma boundary. Both the isothermal (purely toroidal flow) and the poloidal flow cases are treated. Experiment-relevant states with steady flow (so far only in the toroidal direction) are computed by the modified SELENE40 code. (author)
Ion diagnostics for the tokamak boundary
International Nuclear Information System (INIS)
Matthews, G.F.
1991-01-01
In this paper, recent developments in ion diagnostic probes for tokamak boundary plasmas are discussed. Three areas are covered: retarding field analysers, sniffer probes and plasma ion mass spectrometers. The contribution of these diagnostics to our understanding of plasma surface interactions is summarised. (author)
Anomalous periodic disruptions in tokamak plasma
International Nuclear Information System (INIS)
Montvai, A.; Tegze, M.; Valyi, I.
1982-09-01
Anomalously strong, periodic instabilities were observed in the MT-1 tokamak. Characteristics of these instabilities were partly similar to those of internal disruptions, but there were features making them different from the normal relaxational oscillations. Basic characteristics of the phenomenon were studied with the aid of generally used diagnostics. (author)
Energy confinement of high-density tokamaks
Schüller, F.C.; Schram, D.C.; Coppi, B.; Sadowski, W.
1977-01-01
Neoclassical ion heat conduction is the major energy loss mechanism in the center of an ohmically heated high-d. tokamak discharge (n>3 * 1020 m-3). This fixes the mutual dependence of plasma quantities on the axis and leads to scaling laws for the poloidal b and energy confinement time, given the
Supravodivý tokamak dobyl Asii
Czech Academy of Sciences Publication Activity Database
Řípa, Milan
2006-01-01
Roč. 54, č. 18 (2006), s. 58 ISSN 0040-1064 Institutional research plan: CEZ:AV0Z20430508 Keywords : superconducting tokamak * ITER * Tore Supra * Institute of Plasma Physics AV CR Subject RIV: BL - Plasma and Gas Discharge Physics
Runaway electrons in the TRIAM-1 tokamak
International Nuclear Information System (INIS)
Satoh, Takemichi; Nakamura, Kazuo; Toi, Kazuo; Nakamura, Yukio; Hiraki, Naoji
1981-01-01
Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges. (author)
Runaway electrons in the TRIAM-1 tokamak
Energy Technology Data Exchange (ETDEWEB)
Satoh, T; Nakamura, K; Toi, K; Nakamura, Y; Hiraki, N [Kyushu Univ., Fukuoka (Japan). Research Inst. for Applied Mechanics
1981-09-01
Pulse height analysis of soft X-rays is carried out in the TRIAM-1 tokamak. The electron temperatures determined from the soft X-ray spectrum agree well with those from Thomson scattering. It is observed that low-energy runaway (slideaway) electrons appear in the high-current-density discharges.
Plasma position control in SST1 tokamak
Indian Academy of Sciences (India)
also placed inside the vessel, however the controller would ignore fast but insignificant changes in radius arising ... poloidal cross-sectional view of the SST1 plasma along with the stabilizers are shown in figure 1 and ... [1] model which has shown excellent agreement with control experiments in TCV tokamak and also with ...
Central control system for the EAST tokamak
International Nuclear Information System (INIS)
Sun Xiaoyang; Ji Zhenshan; Wu Yicun; Luo Jiarong
2008-01-01
The architecture, the main function and the design scheme of the central control system and the collaboration system of EAST tokamak are described. The main functions of the central control system are to supply a union control interface for all the control, diagnoses, and data acquisition (DAQ) subsystem and it is also designed to synchronize all those subsystem. (authors)
Microinstabilities in weak density gradient tokamak systems
International Nuclear Information System (INIS)
Tang, W.M.; Rewoldt, G.; Chen, L.
1986-04-01
A prominent characteristic of auxiliary-heated tokamak discharges which exhibit improved (''H-mode type'') confinement properties is that their density profiles tend to be much flatter over most of the plasma radius. Depsite this favorable trend, it is emphasized here that, even in the limit of zero density gradient, low-frequency microinstabilities can persist due to the nonzero temperature gradient
Design parameters of Tokamak-7 system
International Nuclear Information System (INIS)
Ivanov, D.P.; Keilin, V.E.; Klimenko, E.Yu.; Strelkov, V.S.
Superconducting windings for the main magnetic field of Tokamak-7 are discussed. The parameters of this facility are based on the use of commercially available superconducting materials for fields up to 80 kOe. Experimental parameters are described. (U.S.)
Modular pulse sequencing in a tokamak system
International Nuclear Information System (INIS)
Chew, A.C.; Lee, S.; Saw, S.H.
1992-01-01
Pulse technique applied in the timing and sequencing of the various part of the MUT tokamak system are discussed. The modular architecture of the pulse generating device highlights the versatile application of the simple physical concepts in precise and complicated research experiment. (author)
INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
International Nuclear Information System (INIS)
HUMPHREYS, D.A.; FERRON, J.R.; JOHNSON, R.D; LEUER, J.A.; PENAFLOR, B.G.; WALKER, M.L.; WELANDER, A.S.; KHAYRUTDINOV, R.R; DOKOUKA, V.; EDGELL, D.H.; FRANSSON, C.M.
2004-03-01
OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance
Tokamak startup with electron cyclotron heating
Energy Technology Data Exchange (ETDEWEB)
Holly, D J; Prager, S C; Shepard, D A; Sprott, J C
1980-04-01
Experiments are described in which the startup voltage in a tokamak is reduced by approx. 60% by the use of a modest amount of electron cyclotron resonance heating power for preionization. A 50% reduction in volt-second requirement and impurity reflux are also observed.