Development of plasma sources for Dipole Research EXperiment (DREX)

Dipole Research EXperiment(DREX) is a new terrella device as part of the Space Plasma Environment Research Facility(SPERF) for laboratory studies of space physics relevant to the inner magnetospheric plasmas. Adequate plasma sources are very important for DREX to achieve its scientific goals. According to different research requirements, there are two density regimes for DREX. The low density regime will be achieved by an electron cyclotron resonance(ECR) system for the ‘whistler/chorus' wave investigation, while the high density regime will be achieved by biased cold cathode discharge for the desired ‘Alfvén' wave study. The parameters of ‘whistler/chorus' waves and ‘Alfvén' waves are determined by the scaling law between space and laboratory plasmas in the current device. In this paper, the initial design of these two plasma sources for DREX is described. Focus is placed on the chosen frequency and operation mode of the ECR system which will produce relatively low density ‘artificial radiation belt' plasmas and the seed electrons, followed by the design of biased cold cathode discharge to generate plasma with high density.     

Conceptual design of the three-dimensional magnetic field configuration relevant to the magnetopause reconnection in the SPERF

A new ground-based experimental device,the Space Plasma Environment Research Facility (SPERF),is being designed at Harbin Institute of Technology in China,with Asymmetric REconnection eXperiment-3 Dimensional (AREX-3D) as one of the experimental components to study the asymmetric reconnection dynamics relevant to the interaction between the interplanetary and magnetospheric plasmas.The asymmetry in the designed magnetic reconnection process not only refers to the distinct plasma parameters designed for the two upstream regions across the current sheet,but also refers to the inhomogeneity in the direction along the current sheet resulting from the designed 3D magnetic field geometry.These two asymmetries are fundamental features of the reconnection process at the Earth's magnetopause.In experiment,the reconnection process is driven by a set of flux cores through coil-currentramp-up from the 'magnetosheath-side' to interact with a dipole magnetic field generated by the Dipole Research EXperiment (DREX) coil on the 'magnetosphere-side'.The AREX-3D will be able to investigate a range of important reconnection issues in 3D magnetic field geometry that is relevant to the Earth's magnetopause.A wide range of plasma parameters can be achieved through inductive plasma generation with flux cores on the 'magnetosheath-side' and electron cyclotron resonance (ECR) with microwave sources on the 'magnetosphere-side',e.g.high (low) plasma density at experimental magnetosheath (dipole) side.Different reconnection regimes and geometries can be produced by adjusting plasma parameters and coil setups as well as coil current waveforms.The three-dimensional magnetic field configurations in the SPERF relevant to the dayside magnetopause reconnection are discussed in detail.     

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40?mm inner diameter coolant pipes providing water for the divertor targets during the “baking” regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. “Baking” regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal–hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1?s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.     

Field Reversed Configuration Translation and the Magnetized Target Fusion Collaboration

After considerable design and construction, we describe the status of a physics exploration of magnetized target fusion (MTF) that will be carried out with the first flux conserving compression of a high pressure field-reversed configuration (FRC). The upgraded Los Alamos (LANL) high density FRC experiment FRXL has demonstrated that an appropriate FRC plasma target can be created and translated on a time scale fast enough to be useful for MTF. Compression to kilovolt temperature is expected to form a Mbar pressure, high energy density laboratory plasma (HEDLP). Integrated hardware on the new Field Reversed Compression and Heating Experiment (FRCHX) at the Air Force Research Laboratory Shiva Star facility, has formed initial FRC’s and will radially compress them within a cylindrically symmetric aluminum “liner”. FRXL has shown that time scales for FRC translation to the target region are significantly shorter than the typical FRC lifetime. The hardware, diagnostics, and design rationales are presented. Pre-compression plasma formation and trapping experimental data from FRXL and FRCHX are shown.     

Improving the NAA laboratory pneumatic transfer system for using Tehran research reactor

Neutron activation analysis (NAA) is a precise multi-elemental method which performs well for qualitative and quantitative analyses. This method has been applied to the laboratory of Radiation Applications Research School (RARS) at University of Tehran. The new pneumatic transfer system (Rabbit) has been designed and constructed for transferring samples, in particular short half-life samples, to Tehran Research Reactor (TRR) core for NAA. With this system samples were transferred and returned to/from the reactor as fast as possible in both automatic and manual modes. This system has two distinct paths for sending and receiving with transfer time of 30 s to the reactor core and 36 s from the irradiation position to the counting station in NAA laboratory which is about 700 m from the laboratory. Experiment and calculation have been carried out for calibration of the neutron flux and spectrum.     

Initial damage processes for diamond film exposure to hydrogen plasma

Diamond is considered to be a possible alternative to other carbon based materials as a plasma facing material in nuclear fusion devices due to its high thermal conductivity and resistance to chemical erosion. In this work CVD diamond films were exposed to hydrogen plasma in the MAGnetized Plasma Interaction Experiment (MAGPIE): a linear plasma device at the Australian National University which simulates plasma conditions relevant to nuclear fusion. Various negative sample stage biases of magnitude less than 500 V were applied to control the energies of impinging ions. Characterisation results from SEM, Raman spectroscopy and ERDA are presented. No measureable quantity of hydrogen retention was observed, this is either due to no incorporation of hydrogen into the diamond structure or due to initial incorporation as a hydrocarbon followed by subsequent etching back into the plasma. A model is presented for the initial stages of diamond erosion in fusion relevant hydrogen plasma that involves chemical erosion of non-diamond material from the surface by hydrogen radicals and damage to the subsurface region from energetic hydrogen ions. These results show that the initial damage processes in this plasma regime are comparable to previous studies of the fundamental processes as reported for less extreme plasma such as in the development of diamond films.     

非能动试验台架中CMT的比例分析及失真评价

堆芯补水箱（CMT）是AP1000非能动堆芯冷却系统中的关键设备,对其进行合理的比例分析对非能动整体性能试验台架的设计起着重要作用。采用H2TS比例分析方法对CMT的循环模式和排水模式进行比例分析,进而将得到的CMT重要过程的相似准则应用于我国正在设计建造的ACME台架的CMT比例设计,并对其特征 Π 群的比例失真度进行定量化计算。最后,对ACME台架的CMT进行比例失真原因分析和评价。结果表明,CMT循环阶段的主要过程能在ACME中得到较好的模拟,而在排水阶段由于ACME超比例的CMT金属质量引起的储冷问题导致蒸汽冷凝过程存在一定的失真,但综合分析认为ACME台架采用高压模拟方案能较好地复现原型电站CMT的重要现象和过程。     

Using the tritium plasma experiment to evaluate ITER PFC safety

The Tritium Plasma Experiment was assembled at Sandia National Laboratories, Livermore and is being moved to the Tritium Systems Test Assembly facility at Los Alamos National Laboratory to investigate interactions between dense plasmas at low energies and plasma-facing component materials. This apparatus has the unique capabilty of replicating plasma conditions in a tokamak divertor with particle flux densities of 2 × 10²³ ions/m².s and a plasma temperature of about 15 eV using a plasma that includes tritium. An experimental program has been initiated using the Tritium Plasma Experiment to examine safety issues related to tritium in plasma-facing components, particularly the ITER divertor. Those issues include tritium retention and release characteristics, tritium permeation rates and transient times to coolant streams, surface modification and erosion by the plasma, the effects of thermal loads and cycling, and particulate production. An industrial consortium led by McDonnell Douglas will design and fabricate the test fixtures.Prepared for the U.S. Department of Energy, Office of Energy Research under DOE Idaho Field Office Contract DE-AC07-76ID01570.     

High heat flux capabilities of the Magnum-PSI linear plasma device

Magnum-PSI is an advanced linear plasma device uniquely capable of producing plasma conditions similar to those expected in the divertor of ITER both steady-state and transients. The machine is designed both for fundamental studies of plasma–surface interactions under high heat and particle fluxes, and as a high-heat flux facility for the tests of plasma-facing components under realistic plasma conditions. To study the effects of transient heat loads on a plasma-facing surface, a novel pulsed plasma source system as well as a high power laser is available. In this article, we will describe the capabilities of Magnum-PSI for high-heat flux tests of plasma-facing materials.     

Analysis of the accident with the coolant discharge into the plasma vessel of the W7-X fusion experimental facility

Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Starting 2007, Lithuanian Energy Institute (LEI) is a member of European Fusion Development Agreement (EFDA) organization. LEI is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) in the frames of EFDA project by performing safety analysis of fusion device W7-X. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. In this paper the safety analysis of 40 mm inner diameter coolant pipe rupture in cooling circuit and discharge of steam–water mixture through the leak into plasma vessel during the W7-X no-plasma “baking” operation mode is presented. For the analysis the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers) and plasma vessel was developed by employing system thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code. This paper demonstrated that the developed RELAP5 model enables to analyze the processes in divertor cooling system and plasma vessel. The results of analysis demonstrated that the proposed burst disc, connecting the plasma vessel with venting system, opens and pressure inside plasma vessel does not exceed the limiting 1.1 × 10⁵ Pa absolute pressure. Thus, the plasma vessel remains intact after loss-of-coolant accident during no-plasma operation of Wendelstein 7-X experimental nuclear fusion facility.     

The design of the Tokamak Physics Experiment (TPX)

The Tokamak Physics Experiment is designed to develop the scientific basis for a compact and continuously operating tokamak fusion reactor. It is based on an emerging class of tokamak operating modes, characterized by beta limits well in excess of the Troyon limit, confinement scaling well in excess of H-mode, and bootstrap current fractions approaching unity. Such modes are attainable through the use of advanced, steady state plasma controls including strong shaping, current profile control, and active particle recycling control. Key design features of the TPX are superconducting toroidal and poloidal field coils; actively-cooled plasma-facing components; a flexible heating and current drive system; and a spacious divertor for flexibility. Substantial deuterium plasma operation is made possible with an in-vessel remote maintenance system, a lowactivation titanium vacuum vessel, and shielding of ex-vessel components. The facility will be constructed as a national project with substantial participation by U.S. industry. Operation will begin with first plasma in the year 2000.     

Varying Electron Cyclotron Resonance Heating on the Levitated Dipole Experiment

Plasmas in the Levitated Dipole Experiment (LDX) are formed and sustained currently via two electron cyclotron resonance heating (ECRH) sources: 2.5 kW at 2.45 GHz and 2.5 kW at 6.4 GHz. An important topic being investigated is how varying the ECRH affects the confinement and stability of the plasma. We report the results of using different operational combinations of our RF sources, such as varying the power composition, changing the power levels, and sequencing of the onset time.     

Equilibrium Reconstruction of Anisotropic Pressure Profile in the Levitated Dipole Experiment

The Levitated Dipole Experiment (LDX) is an innovative confinement concept that uses an internal superconducting dipole field to confine plasma. Plasma equilibrium is calculated by a least-squares fit of an anisotropic pressure model to magnetic measurements constrained by X-ray images. Reconstructions have been done for different heating schemes using two-frequency electron cyclotron heating at 2.45 and 6.4 GHz. Results show that a maximum local β ∼ 20% has been achieved using two frequency heating at a combined full power of 5 kW. Analysis of the reconstruction results shows that the magnetic sensors are sensitive primarily to changes in the plasma dipole moment. This is partly due to the fact that the dipole current decreases as the plasma current increases (as required by flux conservation through the superconducting dipole) and the magnetic sensors detect the sum of these changes. This paper will present details of the reconstruction procedure and describe how new magnetic sensors will aid in resolving the pressure profile more accurately.     

Natural circulation in a liquid metal one-dimensional loop

A wide use of pure lead, as well as its alloys (such as lead-bismuth, lead-lithium), is foreseen in several nuclear-related fields: it is studied as coolant in critical and sub-critical nuclear reactors, as spallation target for neutron generation in several applications and for tritium generation in fusion systems. In this framework, a new facility named NAtural CIrculation Experiment (NACIE), has been designed at ENEA-Brasimone Research Centre. NACIE is a rectangular loop, made by stainless steel pipes. It consists mainly of a cold and hot leg and an expansion tank installed on the top of the loop. A fuel bundle simulator, made by three electrical heaters placed in a triangular lattice, is located in the lower part of the cold leg, while a tube in tube heat exchanger is installed in the upper part of the hot leg. The adopted secondary fluid is THT oil, while the foreseen primary fluid for the tests is lead-bismuth in eutectic composition (LBE). The aim of the facility is to carry out experimental tests of natural circulation and collect data on the heat transfer coefficient (HTC) for heavy liquid metal flowing through rod bundles. The paper is focused on the preliminary estimation of the LBE flow rate along the loop. An analytical methodology has been applied, solving the continuity, momentum and energy transport equations under appropriate hypothesis. Moreover numerical simulations have been performed. The FLUENT 6.2 CFD code has been utilized for the numerical simulations. The main results carried out from the pre-tests simulations are illustrated in the paper, and a comparison with the theoretical estimations is done.     

CARR铱源试验靶件设计和试验验证

为配合中国先进研究堆(CARR)铱源辐照生产项目,设计制造了铱源试验靶件,对试验靶件的设计参数、结构尺寸进行了介绍。在堆外使用专门的传热装置模拟铱源靶件的外部和内部传热工况,测量了用于模拟辐照罐壁面温度和样品温度的传热装置的壁面温度和内部温度,结果验证了热工分析方法是合适的。入堆试验靶件由含有铱片样品的辐照罐和等量发热的模拟罐组成。堆内试验获得的数据综合验证了试验靶件物理热工的分析结果,这个结果可对CARR铱源辐照生产安全评审提供依据且偏于安全。     

Concept and safety studies of an integrated pyroprocess facility

The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocess facilities for the dry processing research of spent fuel for over 10 years since 1997. A hot cell facility, named the Advanced spent fuel Conditioning Process Facility (ACPF), for the demonstration of a head-end part of pyroprocess was developed in November 2005. An inactive demonstration facility for the integrated pyroprocess, named the PyRoprocess Integrated inactive Demonstration (PRIDE) facility, has been developing since 2007. This facility is equipped with several types of pyroprocess equipment, such as a decladding/voloxidation device, a powder mixing device, electro-reduction equipment, electro-refining equipment, electro-winning equipment, Cd distillation equipment, waste salt regeneration & solidification equipment, and a salt distillation equipment. Safety evaluation and safety related systems are also required to inhibit the release of radioactivity into the environment. Accident evaluations and argon flow characteristics are studied for ensuring the expected safety issues. These conceptual and safety studies will be also used to obtain the basic requirements for design of the Engineering Scale Pyroprocess Facility (ESPF) which is the demonstration facility for feasibility of pyroprocess commercialization.     

Preliminary Design of Neutron Dose Measurement Device for Cosmic-Ray

In order to be applied in cosmic ray, a device was designed which uses a spherical moderator and two types of proportional counters. One is the spherical counter which is imbedded at the center of the sphere moderator. This counter is called as an inner detector. The other six counters is the tube counter. Each is located close to the moderator surface and these counters are called as an outer detector.     

Effects of the Hot Electron Interchange Instability on Plasma Confined in a Dipolar Magnetic Field

The Levitated Dipole Experiment (LDX) explores confinement and stability of plasma created within the dipole field of a strong superconducting magnet. During initial experiments, long-pulse, quasi-steady state discharges that last more than 10 s and have peak beta of more than 20% are studied. The plasma is created by multi-frequency electron cyclotron resonance heating (ECRH) at 2.45 and 6.4 GHz. A population of energetic electrons, with mean energies above 50 keV, dominates the plasma pressure. Creation of high pressure, high beta plasma is possible only when intense hot electron interchange (HEI) instabilities are stabilized by sufficient neutral gas fueling. The instabilities resonate with the magnetic drift motion of the energetic electrons and can cause rapid radial transport. Measurements of the electrostatic and magnetic fluctuations of the HEI instability are described along with observations of the instability’s spectral characteristics. Fluctuations of the outer poloidal field induced by the HEI show a rapid evolution of the perturbed pressure profile.      

Dense Plasma Injection Experiment at MCX

We present preliminary results of the High Density Plasma Injection Experiment at the Maryland Centrifugal Experiment (MCX). HyperV Technologies Corp. has designed, built, and installed a prototype coaxial gun to drive rotation in MCX. This gun has been designed to avoid the blow-by instability via a combination of electrode shaping and a tailored plasma armature. An array of diagnostics indicates the gun is capable of plasma jets with a mass of 160 μg at 70 km/s with an average plasma density above 10¹⁵ cm⁻³. Preliminary measurements are underway at MCX to understand the penetration of the plasma jet through the MCX magnetic field and the momentum transfer from the jet to the MCX plasma. Data will be presented for a wide range of MCX field parameters, and the prospects for future injection experiments will be evaluated.     

Advanced instrumented impact testing facility for characterization of dynamic fracture behavior

The convenient and cheap instrumented Charpy impact test used for the evaluation of dynamic fracture toughness suffers many disadvantages and limitations. As an effort to enhance the reliability and accuracy of the instrumented impact testing, a new pendulum type of test apparatus has been designed and constructed at the Technical Research Centre of Finland (VTT).The new tester contains several novel features which help to avoid the drawbacks of the normal designs and which make the impact testing more reliable and appropriate from the fracture mechanics point of view. For that purpose the tester is equipped with a fully computerized data acquisition system and an optical COD measuring device. The latter has been developed and constructed through the co-operation between Materialprüfungsanstalt (MPA) and VTT. In the new testing facility the determination of fracture toughness is based on the application of the optical COD measuring device. In this paper testing equipment and procedures are described and discussed.