After three years of upgrading work, PLS-II (S. Shin, Commissioning of the PLS-II, JINST, January 2013) is now successfully operating. The top-up operation of the 3 GeV linear accelerator had to be delayed because of some challenges encountered, and PLS-II was run in decay mode at the beginning in March 2012. The main difficulties encountered in the top-up operation of PLS-II are different levels between the linear accelerator and the storage ring, the 14 narrow gap in-vacuum undulators in operation, and the full energy injection by 3 GeV linear accelerator. Large vertical emittance and energy jitter of the linac were the major obstacles that called for careful control of injected beam to reduce beam loss in the storage ring during injection. The following measures were taken to resolve these problems: (1) The high resolution Libera BPM (see http://www.i-tech.si) was implemented to measure the beam trajectory and energy. (2) Three slit systems were installed to filter the beam edge. (3) De-Qing circuit was applied to the modulator system to improve the energy stability of injected beam. As a result, the radiation by beam loss during injection is reduced drastically, and the top-up mode has been successfully operating since 19th March 2013. In this paper, we describe the experimental results of the PLS-II top-up operation and the improvement plan.

In this paper, we use a three dimensional particle-in-cell (PIC) code, MAGIC3D, to investigate the operating characteristics of an ultra-compact, high power reltron tube with self-modulation beam and post-acceleration of modulated beam. The simulation model is a full 3-D model because the reltron is intrinsically a three-dimensional problem due to the asymmetrical geometry of the modulating cavity. In the reltron, the mode of operation is pi/2-mode, providing the means to modulate an electron beam in short distances. Simulations show that ultra-compact reltron with a beam voltage of 120 kV and an acceleration voltage of 800 kV generates 38.3 MW high power microwave at 2.887 GHz with an electronic efficiency of 52.7%.

The variations of plasma parameters on the boundary conditions, especially potential, of plasmas were measured in an inductively coupled plasma source developed to generate a hyperthermal neutral beam. Hyperthermal neutrals can be produced by Auger neutralization when ions with low energy are neutralized by impinging on a metal surface called a reflector. However, the reflector is a significant source of ion drain when it is biased to a negative potential. The plasma potential can be negative with respect to the grounded chamber potential while the reflector is negatively biased, but other plasma parameters, namely density and temperature, are not sensitive to the reflector bias. If the electron loss current into the chamber wall is governed by the space charge limited current law, sustainment of the plasma with a negative potential can be explained in terms of the charge balance equations for quasineutrality.

Non-equal length, three section coupled transmission line couplers are proposed. The proposed structure offers increased design flexibility and the compact circuit design capability over the conventional quarterwave coupled lines. The detailed analysis results and design method are presented along with the numerical and the experimental verification.

In this paper, we fabricate RF integrated passive devices (RFIPDs) on two types of substrate. One is smart silicon substrate with a 25-mu m-thick SiO2 surface, and the other is glass substrate. A low-cost manufacturing technology, BCB interlayer, and 10-mu m Cu plating process are used to implement high-performance RFIPDs. We investigate the power-handling performance with these two types of RFIPDs. Generally, RFIPDs are used for RF front-end module components, so handling power is very important factor. But, these two types of RFIPDs show different handling-power performances. in contrast to the 5.5-W power handling of the LPF on smart silicon substrate, that of the LPF oil glass substrate is below 1.5 W (with CW power and 168 h operating time). (c) 2005 Wiley Periodicals, Inc.

HANBIT is a magnetic mirror confinement device. Recent physics experiments have been mainly focused on identifying discharge characteristics and on getting stable plasma production and operation modes, by using a 500 kW slot antenna system at a fixed RF frequency of 3.5 MHz and varying discharge conditions such as fueling rate, RF power, and B-field intensity in central cell. Two distinct operation modes are found to be very sensitive to the RF power as well as the ratio of the RF frequency to the ion cyclotron frequency. A MHD interchange mode of m =-1 is clearly observed and the stabilizing mechanism can be explained by the sideband coupling theory.

Understanding of electromagnetic coupling is an important area of research in the design and development of multi-layered antenna arrays. In this paper, the spectral domain moment method (SDMM) is developed to investigate the coupling between two microstrip lines through a circular waveguide in a thick common-ground plate. Measured data, along with the results obtained from the finite-difference time-domain (FDTD) method, are used to verify the solution. A parametric study illustrates the coupling behavior of the structure under investigation. It is observed that the frequency tuning and wide frequency bandwidth can be achieved by choosing the appropriate geometry, and this type of transition can be significantly affected by structural parameters. (C) 2003 Wiley Periodicals.

HANBIT is a magnetic mirror confinement device. Recently, after finishing the first campaign for the basic system development, it started the second campaign for high-temperature plasma confinement physics study in a mirror configuration. Here, we introduce briefly the HANBIT device and report initial physics experiment results on RF-plasma heating and confinement in the simple mirror configuration. It appears that the discharge characteristics of HANBIT are quite different from those in other mirror devices, and an explanation is presented to clarify the difference.

Two types of neutral beam sources have been developed in order to measure plasma parameters on the Hanbit mirror device. The first source is a diagnostic neutral beam (DNB) which consists of a hydrogen neutral beam with a beam energy of 30 keV and a total beam current of ca. 1 A. The ion temperature profile can be determined by measuring directly the broadening of the Hα line emitted from hydrogen neutrals produced through the charge exchange recombination reaction with the DNB in the plasma. A fibre optic array detector, which works as an ideal notching filter, was developed to filter out the intense Hα line emitted from the cold hydrogen atoms in the plasma edge. The second source is a hyperthermal neutral beam (HNB) which consists of neutral particles with an energy of 1-100 eV. The HNB can be used to measure electron temperature and density profiles in the region between the core and the outer edge. This region cannot be covered either by Thomson scattering or by electrostatic probes. The feasibility of obtaining profiles of electron density and temperature by means of a helium HNB with a collisonal radiative equilibrium code has been performed.

The first phase of the HANBIT project that started in 1996, is ending this year. The goal of the first phase is to establish a decent experimental facility based on the HANBIT mirror device and to set up a framework for operating this machine as a National Joint Users’ Facility. A five-year plan has been being undertaken to improve the performance of HANBIT, refurbished from the old TARA machine, by adding diagnostics and heating capacity. As the goal of the first phase has been achieved, the second phase is being planned, in which basic mirror confinement physics will be pursued as well as plasma applications and training of people. Most of new works are possible via close collaboration with external experts. We are exercising a new form of collaboration between universities and the institute and between industries and the institute through the HANBIT User Program, which is a key ingredient for successful operation of a National Users’ facility like the HANBIT device.

The extensive design effort for KSTAR has been focused on two major aspects of the KSTAR project mission - steady-state-operation capability and advanced tokamak physics. The steady state aspect of the mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long pulse current drive and heating systems. The advanced tokamak aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double null divertor and passive stabilizers, internal control coils and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, the vacuum vessel, plasma facing components and power supplies. The new KSTAR experimental facility with cryogenic system and deionized water cooling and main power systems has been designed, and the construction work is under way for completion in 2004.

The Korea Superconducting Tokamak Advanced Research (KSTAR) project is the major effort of the national fusion programme of the Republic of Korea. Its aim is to develop a steady state capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. The major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T and plasma current 2 MA, with a strongly shaped plasma cross-section and double null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beams, ion cyclotron waves, lower hybrid waves and electron cyclotron waves for flexible profile control in advanced tokamak operating modes. A comprehensive set of diagnostics is planned for plasma control, performance evaluation and physics understanding. The project has completed its conceptual design and moved to the engineering design and construction phase. The target date for the first plasma is 2002.

The five ARIES designs, which correspond to five different tokamak operating modes, are reviewed and compared. Physics figures of merit are introduced that quantify the major parameters of a tokamak design in a physics operating space. The five operating modes are compared to one another and to the existing tokamak data base in terms of these physics parameters. While the steady-state first stability design [ARIES-I] and the pulsed first stability design [PULSAR] are closest, no design has yet been completely prototyped in existing tokamaks.

An advanced heat removal scenario is required to handle the high input power of magnetic fusion reactors. The concepts of gas target and radiative divertors have been explored. A set of self-consistently coupled fluid equations for plasma and neutrals is suitable for this system because of the sophisticated equations for neutrals. The transfer of plasma momentum to neutrals, which is crucial in understanding the plasma-neutral interface, is described through the charge exchange process. Our code results show good agreement with gas target experiments on the PISCES experiment and demonstrate its superiority to the previous simulation. This benchmarked code and a set of equations have been applied to a slot divertor for DEMO reactor. The slot length for the divertor can be determined by producing enough frictional force to overcome the thermal force which creates a backflow of impurities. A small amount of radiating argon is enough to exhaust 40 MW of power from the core plasma in a 40 cm long slot for the inner divertor of DEMO.

The Korea Superconducting Tokamak Advanced Research (KSTAR) Project is the major effort of the Korean National Fusion Program (KNFP) to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor. Major parameters of the tokamak are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 Tesla, and plasma current 2 mA with a strongly shaped plasma cross-section and double-null divertor. The initial pulse length provided by the poloidal magnet system is 20 s, but the pulse length can be increased to 300 s through non-inductive current drive. The plasma heating and current drive system consists of neutral beam, ion cyclotron waves, lower hybrid waves, and electron–cyclotron waves for flexible profile control. A comprehensive set of diagnostics is planned for plasma control and performance evaluation and physics understanding. The project has completed its conceptual design phase and moved to the engineering design phase. The target date of the first plasma is set for year 2002.

The Korean Superconducting Tokamak Advanced Research (KSTAR) will have superconducting magnets for both the poloidal field (PF) coils and the toroidal field (TF) coils. The physical arrangement of the PF set has 14 coils external to the TP coils. The analysis of the equilibrium flexibility of the KSTAR PF system determines the coil currents required to maintain prescribed equilibrium configurations over the desired range of operational parameters specified in I/sub p/ (q/sub 95/), /spl beta//sub N/, and l/sub 1/(3). Constraints on the plasma separatrix and the flux linkage through the geometric center of the plasma are specified for the free boundary equilibrium calculations. In order to eliminate the remaining freedoms the coil current distribution is regularized (smoothed) by minimizing the quantity /sup ncoil//spl Sigma//sub j=1/ (i/sub j///spl Delta/Aj)/sup 2/ where /spl Delta/A/sub j/ is the area of each PF coil. The ripple magnitude due to the finite number of TF coils and the size of the port for the neutral beam injector determine the number, size, rand shape of TF coils. Two ripple criteria for a shaped plasma are used for types of ripple transport. The current design of the TP coil, 16 coils and D-shape, is big enough to satisfy requirements on the ripple magnitude at the plasma and to provide an adequate access for tangential NB injection.

The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets are superconducting coils. The device is configured to be initially capable of 20 s pulse operation and then to be upgraded for operation up to 300 s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radiation shielding.

The reduction of peak heat fluxes on KSTAR (Korea Superconducting Tokamak Advanced Research) divertor plates is required to meet the requirements of <3.5 MW/m(2) heat flux and average and peak tile temperatures of < 600 degrees C and <1200 degrees C, respectively, since numerical simulations give peak heat fluxes of greater than or equal to 4.5 MW/m(2) for the initial heating power of 15.5 MW with no impurity. The radiative divertor simulation obtained with neon concentration of 1.85% could collapse electron temperature and target plate heat flux well below the requirement. Zeff at Core boundary was similar to 3. Study on geometry optimization shows that 7cm of outboard pumping gap gives the best gas throughput and gas throughput can be improved by the proper selection of pumping shape. Core ionization could be reduced by a factor of 2 with inner bump structure. The effects of inboard pumping gap and cryopump are negligible compared to the outboard ones.

The effect of the wall location on the stability of magnetohydrodynamic equilibria is considered for the case of an advanced tokamak with negative central magnetic shear (NCS). The equilibria have pressure and current profiles consistent with the proposed second stable core VH (SSC-VH) mode. In particular, the pressure probe has a finite edge pressure gradient that is consistent with high confinement as seen in experiments. The stability analysis shows that, for q0 = 3.9, all qmin in the range 1.8 ⩽ qmin ⩽ 3.3 are stable to n=1 and n=2 modes at βN = 5.0 with the DIII-D wall. However, there are two optimum minimum q (qmin) with respect to the wall distance from the plasma. For n=1 modes with q0 = 3.9, the optimum qmin is 2.1, with a secondary optimum at qmin = 2.85. The wall position for the n = 2 mode is more restrictive than that for the n = 1 mode but the optimum qmin values for both n = 1 and n = 2 are almost identical

General plasma system requirements for the U.S. Demo fusion power plant have been identified, taking into account factors of economics, operability and reliability required extrapolations from the present plasma performance database have been assessed for the Demo options in five areas of tokamak operation, namely, stability, current drive, energy confinement, heat exhaust, and ash removal, for which figures of merit have been identified