Engineering design for the HL-2M tokamak components

The HL-2A tokamak will be modified into HL-2M. The Bt at the plasma center (major radius R = 1.78 m) is 2.2 T, the minor radius is 0.65 m. The plasma current I_P of HL-2M will reach up to 2.5 MA, the elongation and triangularity is more than 1.8 and more than 0.5, respectively. The vacuum vessel torus consists of 20 sectors with “D” shaped cross-section and double wall structure. 20 toroidal field coil bundles comprise 140 turns which are designed with demountable joints, the poloidal field coils system consists of 25 coils. The engineering design and calculation for field coil system, vacuum vessel, support structure, etc. are finished, many key issues for manufacture process have been discussed with industry and the fabrication of main components of HL-2M tokamak will be carried out in factories.     

Detailed design of the RF source for the 1 MV neutral beam test facility

In the framework of the EU activities for the development of the Neutral Beam Injector for ITER, the detailed design of the Radio Frequency (RF) driven negative ion source to be installed in the 1 MV ITER Neutral Beam Test Facility (NBTF) has been carried out.Results coming from ongoing R&D on IPP test beds [A. Stäbler et al., Development of a RF-Driven Ion Source for the ITER NBI System, this conference] and the design of the new ELISE facility [B. Heinemann et al., Design of the Half-Size ITER Neutral Beam Source Test Facility ELISE, this conference] brought several modifications to the solution based on the previous design.An assessment was carried out regarding the Back-Streaming positive Ions (BSI+) that impinge on the back plates of the ion source and cause high and localized heat loads. This led to the redesign of most heated components to increase cooling, and to different choices for the plasma facing materials to reduce the effects of sputtering.The design of the electric circuit, gas supply and the other auxiliary systems has been optimized. Integration with other components of the beam source has been revised, with regards to the interfaces with the supporting structure, the plasma grid and the flexible connections.In the paper the design will be presented in detail, as well as the results of the analyses performed for the thermo-mechanical verification of the components.     

Conceptual design of Remote Control System for EAST tokamak

The international collaboration becomes popular in tokamak research like in many other fields of science, because the experiment facilities become larger and more expensive. The traditional On-site collaboration Model that has to spend much money and time on international travel is not fit for the more frequent international collaboration. The Remote Control System (RCS), as an extension of the Central Control System for the EAST tokamak, is designed to provide an efficient and economical way to international collaboration. As a remote user interface, the RCS must integrate with the Central Control System for EAST tokamak to perform discharge control function. This paper presents a design concept delineating a few key technical issues and addressing all significant details in the system architecture design. With the aim of satisfying system requirements, the RCS will select rich Internet application (RIA) as a user interface, Java as a back-end service and Secure Socket Layer Virtual Private Network (SSL VPN) for securable Internet communication.     

Design of the HL-2M power supply control system

The power supply control system is one of the most important local control systems of HL-2M tokamak. The power supply system of HL-2M is composed of four flywheel generator sets with total capacity of 600 MVA, all the magnetic field coil power supplies and high voltage power supplies of auxiliary heating system. The control system of power supply should make sure the large amount of on-site equipment and facilities to operate steadily and reliably.This paper presents the architecture of HL-2M control system and four kinds of established control networks. It describes how the power supply control system is built by using the established hardware structures and software agreements. Specially, it is focused on introducing an application of real-time technology based on the reflective shared memory and the fully digital compact solution for controlling the high power converters. Otherwise, a scheme which adopts PROFINET and Sequence of Events technology to carry out the intellect interlock control system is given.     

Characteristics of modified CT injector for JFT-2M

The HIT-CTI mark II compact toroid (CT) injector employed for the JFT-2M tokamak facility at the Japan Atomic Energy Research Institute (JAERI) has been upgraded to improve injection performance. The nozzle of the mark III injector now has a linear tube in place of the original focus cone to avoid rapid focus and deceleration, and the tapered outer electrode has been replaced with more gentle taper in the compression section in order to facilitate gradual compression. The dependence of CT velocity and electron density on poloidal bias flux and trigger time of CT acceleration have been investigated in the operable range of 70–230 km/s average CT velocity and electron density of 0.1–1.0 × 10²² m⁻³ at an accelerator bank voltage of 25 kV. The operation window is broader than that of the mark II injector. Emission of a CT plasmoid from the injector, and transport to the flux conserver as a high-density spheromak magnetic structure have also been confirmed.     

NBI双潘宁离子源参数的静电探针测量

介绍了静电探针在22cm双潘宁离子源中心等离子体参数测量中的应用.测出了探针的伏安特性曲线,由该曲线求出在弧流为200 A时,电子温度约为6.0 eV,密度约为2×1012 cm-3,这对于调整放电控制条件,获得高效率大功率的离子源有重要的参考价值,并且对于今后实现强而且均匀的中性束束流的引出也有着重要意义.     

Magnetic configuration flexibility of snowflake divertor for HL-2M

HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reduce the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. A new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.     

Criticality in the fabrication of ion extraction system for SST-1 neutral beam injector

For the heating of plasma in steady-state superconducting tokamak (SST-1) (Y.C. Saxena, SST-1 Team, Present status of the SST-1 project, Nucl. Fusion 40 (2000) 1069–1082; D. Bora, SST-1 Team, Test results on systems developed for the SST-1 tokamak, Nucl. Fusion 43 (2003) 1748–1758), a neutral beam injector is provided to raise the ion temperature to 1 keV. This injector has a capability of injecting hydrogen beam with the power of 0.5 MW at 30 keV. For the upgrade of SST-1, power of 1.7 MW at 55 KeV is required. Further, beam power is to be provided for a pulse length of 1000S. We have designed a neutral beam injector (S.K. Mattoo, A.K. Chakraborty, U.K. Baruah, P.K. Jayakumar, M. Bandyopadhyay, N. Bisai, Ch. Chakrapani, M.R. Jana, R. Onali, V. Prahlad, P.J. Patel, G.B. Patel, B. Prajapati, N.V.M. Rao, S. Rambabu, C. Rotti, S.K. Sharma, S. Shah, V. Sharma, M.J. Singh, Engineering design of the steady-state neutral beam injector for SST-1, Fusion Eng. Des. 56 (2001) 685–691; A.K. Chakraborty, N. Bisai, M.R. Jana, P.K. Jayakumar, U.K. Baruah, P.J. Patel, K. Rajasekar, S.K. Mattoo, Neutral beam injector for steady-state superconducting tokamak, Fusion Technol. (1996) 657–660; P.K. Jayakumar, M.R. Jana, N. Bisai, M. Bajpai, N.P. Singh, U.K. Baruah, A.K. Chakraborty, M. Bandyopadhyay, C. Chrakrapani, D. Patel, G.B. Patel, P. Patel, V. Prahlad, N.V.M. Rao, C. Rotti, V. Sreedhar, S.K. Mattoo, Engineering issues of a 1000S neutral beam ion source, Fusion Technol. 1 (1998) 419–422) satisfying the requirements for both SST-1 and its upgrade. Since intense power is to be transported to SST-1 situated at a distance of several meters from the ion source, the optical quality of the beam becomes a primary concern. This in turn, is determined by the uniformity of the ion source plasma and the extractor geometry. To obtain the desired optical quality of the beam, stringent tolerances are to be met during the fabrication of ion extractor system.

SST-1 neutral beam injector is based on positive ion source. The extraction system consists of three grids, each having extraction area of (width) 230 mm × (height) 480 mm and 774-shaped apertures of 8-mm diameter. To obtain horizontal focal length of 5.4 m and vertical of 7 m, each grid consists of two halves with 387 apertures. Two halves are inclined at an angle of 1.07 ± 0.01°. For long pulse operation, active water cooling is provided by in-laid down of dense network of 22 wavy semicircular (r = 1.1 ± 0.05 mm) cooling channels in the space available between the apertures. The required flatness of the copper plate is 100 μm and positioning tolerance of aperture is ±60 μm. The measurement obtained after fabrication is compared with the specifications. It is pointed out that fabrication within set tolerance limit could be achieved only through process of fabrication and high-resolution measurements.     

Comparison of different plasma chambers in microwave ion source for the intense neutron tube

The microwave absorption efficiency, which is relevant to magnet field and its distribution, is a major parameter of the microwave ion source (MWIS) for the intense neutron tube. Based on previous work, the relations between microwave absorption efficiency and plasma chamber structure and thickness of the microwave introduction window are studied. The microwave absorption efficiency reaches to 100% when plasma chamber is 100mm long and the window thickness is 30mm. The microwave absorption efficiency as a function of pressure is also presented.     

Conceptual Design of 25-Barrel Pellet Injector with Cycle Refrigerator for HL-2A Tokamak

The preliminary design of a multi-barrels pellet injector with cycle refrigerator as an advanced plasma-fuelling tool for HL-2A tokamak has been proposed. The design aims at precise temperature control, easy operation with high reliability and high flexibility. GM-cycle refrigerator and pipe-gun structure have been employed to produce 25 pellets in 25 gun barrels simultaneously and the design aims. have been accomplished. Prime design principle, engineering parameters, structure and layout of the cryostat components as well as calculation of heat load for the cryostat are presented.     

Design of the Eelectron Cyclotron Resonance Heating Transmission System for HL-2A

HL-2A will be equipped with a 75 GHz/1 MW/ls electron cyclotron resonance heating (ECRH) system. The paper describes the design of the transmission system, which is made up of a transmission line, an equatorial launcher and a measureing system. The paper describes in detail the design of main components of the system such as, waveguides, ellipsoidal surface mirrors, chemical vapor (CVD) diamond window, steering mirror, and new directional coupler with a k-spectrometer. The newly-designed launcher can make the beam steer poloidally to deposit energy in different locations.     

Evaluation of a numeric procedure for flow simulation of a 5 × 5 PWR rod bundle with a mixing vane spacer

The fuel assemblies of the Pressurized Water Reactors (PWR) are constituted of rod bundles arranged in a regular square configuration by spacer grids placed along its length. The presence of the spacer grids promote two antagonist effects on the core: a desirable increase of the local heat transfer downstream the grids and an adverse increase of the pressure drop due to the constriction on the coolant flow area. Most spacer grids are designed with mixing vanes which cause a cross and swirl flow between and within the subchannels, enhancing even more the heat transfer performance in the grid vicinity. The improvement of the heat transfer increases the departure from the nucleate boiling ratio, allowing higher operating power in the reactor. Due to these important thermal and fluid dynamic features, experimental and theoretical investigations have been carried out in the past years for the development of spacer grid design. More recently, the Computational Fluid Dynamics (CFD) using three dimensional Reynolds Averaged Navier Stokes (RANS) analysis has been used efficiently for this purpose. Many computational works have been performed, but the appropriate numerical procedure for the flow in rod bundle simulations is not yet a consensus. This work presents results of flow simulations performed with the commercial code CFX 11.0 in a PWR 5 × 5 rod bundle segment with a split vane spacer grid. The geometrical configuration and flow conditions used in the experimental studies performed by Karoutas et al. were assumed in the simulations. To make the simulation possible with a limited computational capacity and acceptable mesh refinement, the computational domain was divided in 7 sub-domains. The sub-domains were simulated sequentially applying the outlet results of a previous sub-domain as inlet condition for the next. In this study the k-ε turbulence model was used. The simulations were also compared with those performed by Karoutas et al. in half a subchannel and In et al. in one subchannel computational domains. Comparison between numerical and experimental results of lateral and axial velocities along of the rod bundle show good agreement for all evaluated heights downstream the spacer grid. The present numerical procedure shows better predictions than Karoutas et al. model especially further from the spacer grid where the peripheral subchannels have more influence in the average flow.     

Application of DEGAS for Ion Temperature Profile Reconstruction from a NPA Diagnostic on HL-2A

On HL-2A, the ion temperature profile can be roughly estimated by analyzing the energy spectrum of the neutral particles that escape from core plasma through 11 different spatial channels. However, this profile deviates from the real radial distribution because of the averaging and attenuation effects. The DEGAS code is used to estimate these effects and reconstruct a more reasonable ion temperature profile by simulating these neutral spectra and making them conform to the measurements. In this paper, critical parameters that affect the escaped neutral spectra are enumerated, and an example of a reconstruction of the ion temperature profile is presented based on experimental data.     

HL-2A装置水平场电源系统局部仿真(阶段报告)

根据中国环流器 ?2 号(HL-2A)装置对水平场电源提出的苛刻要求,设计了一种晶体管数字变流器供电方式,并用电路分析软件PSPICE 对此供电方式进行了数字电路仿真。结果表明,这种晶体管数字变流器在响应速度及波形质量等方面均能满足实验装置的要求。     

Experimental and calculation results of the integral reflood test QUENCH-14 with M5® cladding tubes

The QUENCH-14 experiment investigated the effect of M5® cladding material on bundle oxidation and core reflood, in comparison with tests QUENCH-06 (ISP-45) that used standard Zircaloy-4 and QUENCH-12 that used VVER E110-claddings. The PWR bundle configuration of QUENCH-14 with a single unheated rod, 20 heated rods, and four corner rods was otherwise identical to QUENCH-06. The test was conducted in principle with the same protocol as QUENCH-06, so that the effects of the change of cladding material could be observed more easily. Pre-test calculations were performed by the Paul Scherrer Institut (Switzerland) using the SCDAPSIM, SCDAP/RELAP5 and MELCOR codes. Follow-on post-test analyses were performed using SCDAP/RELAP5 and MELCOR as part of an ongoing programme of model validation and code assessment. Alternative oxidation correlations were used to examine the possible influence of the M5® cladding material on hydrogen generation, in comparison with Zircaloy-4.     

Preliminary design of the control and data acquisition systems for the Neutral Beam Test Facility

The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600 s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC.     

1-D thermal-hydraulic analysis of the high temperature superconducting current leads for the ITER magnet system from 5 K to 300 K

The magnet system of ITER includes high temperature superconducting (HTS) current leads with a maximum current of 68 kA for the toroidal field (TF) coils, 55 kA for the poloidal field (PF)/central solenoid (CS) coils and 10 kA for the control coils (CC), respectively. Although different in terms of size and operative conditions, the ITER HTS current leads have been all designed on the basis of an established concept, which was successfully developed for the LHC at CERN and proven by the so-called 70 kA “demonstrator” lead made by KIT and by the ITER pre-prototypes made by ASIPP in China. A broad R&D campaign has been undertaken by ASIPP and CERN in order to find optimized designs for each component of the leads. Nevertheless, a comprehensive picture of the performance of the entire HTS current leads is not yet available. In this paper, a steady state, full length, thermal-hydraulic 1-D modeling is applied to the study of the three types (TF, PF/CS, CC) of ITER HTS current leads. The results of this predictive analysis are then compared with relevant ITER requirements. It was found that the present design of the HTS current leads will fulfill these specifications.     

Reactor core design optimization of the 200 MWt Pb–Bi cooled fast reactor for hydrogen production

In this study reactor core geometrical optimization of 200 MWt Pb–Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540 °C. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550 °C and the maximum coolant outlet temperature less than 700 °C. By taking into account of the hydrogen production as well as corrosion resulting from Pb–Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350 °C and the coolant flow rate of 7000 kg/s were preferred as the best design parameters.     

Development of micro-beam NRA for 3D-mapping of hydrogen distribution in solids: Application of tapered glass capillary to 6 MeV N ion

A micro-beam NRA system, by means of a resonant nuclear reaction ¹H(¹⁵N, αγ)¹²C, has been developed for the purpose of the 3D mapping of the hydrogen distribution in solids. To obtain the tens μm size of the beam spot, the combination of the newly proposed tapered glass capillary and a conventional quadrupole magnetic lens is employed. An Y patterned film on a substrate is prepared as an application of the developed system. The 6 MeV ¹⁵N beam focused by glass capillaries down to 50 μm successfully shows the hydrogen distribution. The in-plane NRA profile implies that the beam emitted from the glass capillary outlet is parallel, although the original beam has a considerable divergence. The NRA measurements in the 10³ Pa N₂ atmosphere due to the low gas conductance of the glass capillary is also demonstrated.