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Karamyshev O.V.,JINR | Karamysheva G.A.,JINR | Morozov N.A.,JINR | Samsonov E.V.,JINR | And 6 more authors.
IPAC 2016 - Proceedings of the 7th International Particle Accelerator Conference | Year: 2016

The SC200 superconducting cyclotron for hadron therapy is under development by collaboration of ASIPP (Hefei, China) and JINR (Dubna, Russia). The accelerator will provide 200 MeV proton beam with maximum current of 1μA in 2017-2018. The cyclotron is very compact and light, the estimate total weight is about 35 tons and extraction radius is 60 cm. We have performed simulations of all systems of the SC200 cyclotron and specified the main parameters of the accelerator. Average magnetic field of the cyclotron is up to 3.5 T and the particle revolution frequency is about 45 MHz, these parameters increases the requirements for accuracy of the beam dynamics studies. We have designed and performed beam tracking starting from the ion source. Codes and methods used for the beam tracking are presented. Copyright © 2016 CC-BY-3.0 and by the respective authors.

Dr. Charles B. Daknis, Co-Founder, Spine & Pain Centers of New Jersey & New York, has joined The Expert Network, an invitation-only service for distinguished professionals. Dr. Daknis has been chosen as a Distinguished Doctor™ based on peer reviews and ratings, numerous recognitions, and accomplishments achieved throughout his career. Dr. Daknis outshines others in his field due to his extensive educational background, numerous awards and recognitions, and career longevity. After graduating summa cum laude from St. Joseph’s University in Philadelphia, Dr. Daknis graduated at the top of his class at Drexel University’s Hahnemann School of Medicine. His drive and passion for medicine were evident from early on in is training. At Drexel, he received the Nelson Neurosurgical award for excellence in the study of Neurosciences and was appointed as a delegate to the National Institutes of Health for Neuroscience research. Dr. Daknis earned his medical degree in 1990 and went on to complete a residency in Anesthesia at the Albany Medical Center University Hospital where he served as Chief Resident. He followed with a fellowship at Harvard University’s world-renowned Brigham and Women’s Hospital in Boston. Dr. Daknis is quadruple board certified. With over 20 years dedicated to medicine, Dr. Daknis brings a wealth of knowledge to his industry, and in particular, to his areas of expertise, interventional pain management and spinal surgery. When asked why he decided to pursue a career in this specialization, Dr. Daknis said: "I remember being in the second grade and loving every time I went to see my Doctor. I thought he was so professional and knowledgeable. He always helped me when I was in need. From then on, I always wanted to help people like he did." Over the course of his career, Dr. Daknis has cultivated a reputation as one of the top pain management physicians specializing in spinal pain in the United States. Rated as a Top Doctor by Castle Connolly each year since 2013, he is one of only a few doctors to have board certification in Interventional Pain Management and not just Medical Pain Management. He was also awarded NJ Monthy’s Top Doctor in 2012, 2013, 2014, 2015 and 2016. As a thought-leader in his specialty, Dr. Daknis is a published author in multiple peer-reviewed journals and has authored book chapters on pain management and the treatment of various painful disorders. His research focuses on cervical spine and neck disorders and brings this focus and passion for learning to his community. Dr. Daknis speaks regularly on the topic of spine and pain, including a speaking engagement at the McGuire Airforce base at Fort Dix Lakehurst to discuss spinal cord stimulation with the troops. Additionally, he is a leading voice in spinal education and an active board member of the North American Spine Society, the Spine Intervention Society (SIS), as well as the Ethics Committee for the American Society of Interventional Pain Physicians (ASIPP). He is also on the Evidence Analysis Committee for Spine Intervention Society. This prominence in his field gives Dr. Daknis a unique vantage point from which to follow recent developments in technology in pain management. To him, staying on top of the newest techniques and advances is the key to best caring for his patients. He noted: "We look at a lot of stem cell technology and believe that this will become the norm in the next ten years. Biologic therapy may completely transform the possibilities in the world of spine care and pain, as well as regenerative therapies to restore disc health. I believe we may look back at our surgical techniques of today and scoff compared to the realm of opportunity ahead." The Expert Network© has written this news release with approval and/or contributions from Dr. Charles B. Daknis. The Expert Network© is an invitation-only reputation management service that is dedicated to helping professionals stand out, network, and gain a competitive edge. The Expert Network selects a limited number of professionals based on their individual recognitions and history of personal excellence.

Pautasso G.,Max Planck Institute for Plasma Physics (Garching) | Zhang Y.,ASIPP | Reiter B.,Max Planck Institute for Plasma Physics (Garching) | Giannone L.,Max Planck Institute for Plasma Physics (Garching) | And 11 more authors.
Nuclear Fusion | Year: 2011

This paper describes the most recent contributions of ASDEX Upgrade to ITER in the field of disruption studies. (1) The ITER specifications for the halo current magnitude are based on data collected from several tokamaks and summarized in the plot of the toroidal peaking factor versus the maximum halo current fraction. Even if the maximum halo current in ASDEX Upgrade reaches 50% of the plasma current, the duration of this maximum lasts a fraction of a ms. (2) Long-lasting asymmetries of the halo current are rare and do not give rise to a large asymmetric component of the mechanical forces on the machine. Differently from JET, these asymmetries are neither locked nor exhibit a stationary harmonic structure. (3) Recent work on disruption prediction has concentrated on the search for a simple function of the most relevant plasma parameters, which is able to discriminate between the safe and pre-disruption phases of a discharge. For this purpose, the disruptions of the last four years have been classified into groups and then discriminant analysis is used to select the most significant variables and to derive the discriminant function. (4) The attainment of the critical density for the collisional suppression of the runaway electrons seems to be technically and physically possible on our medium size tokamak. The CO2 interferometer and the AXUV diagnostic provide information on the highly 3D impurity transport process during the whole plasma quench. © 2011 IAEA, Vienna.

Breschi M.,University of Bologna | Devred A.,ITER Organization | Casali M.,University of Bologna | Bessette D.,ITER Organization | And 12 more authors.
Superconductor Science and Technology | Year: 2012

The performance of the toroidal field (TF) magnet conductors for the ITER machine are qualified by a short full-size sample (4m) current sharing temperature (T cs) test in the SULTAN facility at CRPP in Villigen, Switzerland, using the operating current of 68kA and the design peak field of 11.8T. Several samples, including at least one from each of the six ITER Domestic Agencies participating in TF conductor fabrication (China, European Union, Japan, Russia, South Korea and the United States), have been qualified by the ITER Organization after achieving T cs values of 6.06.9K, after 7001000 electromagnetic cycles. These T cs values exceed the ITER specification and enabled the industrial production of these long-lead items for the ITER tokamak to begin in each Domestic Agency. Some of these samples did not pass the qualification test. In this paper, we summarize the performance of the qualified samples, analyze the effect of strand performance on conductor performance, and discuss the details of the test results. © 2012 IOP Publishing Ltd.

Ivanov D.P.,RAS Research Center Kurchatov Institute | Anashkin I.O.,RAS Research Center Kurchatov Institute | Khvostenko P.P.,RAS Research Center Kurchatov Institute | Kolbasov B.N.,RAS Research Center Kurchatov Institute | And 8 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2012

The latest superconducting magnets (SM) for fusion are mostly force-cooled, mainly because it allows reliable electrical insulation of the coils using vacuum pressure impregnation (VPI). SM of this type have many leads, feeders and coolant tubes, located in cryostat vacuum, which must sustain high voltages, induced on them by fast current changes. However vacuum loss can spoil their insulation. A few such cases occurred during the T-15 tokamak coils testing, initially having bare leads relying upon vacuum. But its loss generated a coil quench, a protecting current dump at high voltage, followed by breakdown and arc. Even leads insulation by Teflon and fiberglass tape wrap proved to be insufficient. Nevertheless, similar tape wrap insulation of leads and feeders (ILF) was used in EAST, KSTAR, SST-1 and W-7X. So far, seven breakdowns occurred during their coil tests at operating voltage ∼<3 kV. Breakdowns never initiated in the coils, but always on their leads, feeders and sensor lines, indicating that their insulation made by tape wrap were too weak. Instead of ILF improvement some projects undertake Paschen tests. These are planned as the baseline for ITER too. But these tests are valid for the coil with open insulated surface, but are not appropriate for the final tests, when insulation should not be exposed to vacuum. Up to now ILF final tests have been done in all devices at 10-21 kV, but only in good vacuum in spite of the fact that such tests could not guarantee safe operation in case of vacuum loss. We propose to increase ILF strength to the same level, as in the coils, using vacuum-tight grounded stainless steel casings filled up by VPI over magnet leads. This will provide reliable and easily testable solid insulation. Besides, casings would exclude He leaks, providing the second vacuum tight barrier over the ILF. Thus it would increase the magnet reliability and would make it possible to avoid the needs of all single coils test. © 2011 IEEE.

Ballarino A.,CERN | Bauer P.,ITER Organization | Bi Y.,ASIPP | Devred A.,ITER Organization | And 9 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2012

Following the design, fabrication and test of a series of trial leads, designs of the three types of current leads required for ITER have been developed, and targeted trials of specific features are in progress on the way to fabrication and testing of prototype units. These leads are of the hybrid type with a cold section based on the use of high temperature superconductor (HTS) and a resistive section cooled by forced flow of helium gas, optimized for operation at 68 kA, 55 kA and 10 kA. The leads incorporate relevant features of the large series of current leads developed and constructed for the CERN-LHC, relevant features of the trial leads built for ITER, and additional features required to fully satisfy the exigent constraints of ITER with regard to cooling, insulation, and interfaces to feeder and powering systems. In this report a description of the design of the leads is presented, together with plans for the preparation of prototype manufacture and testing at ASIPP. © 2011 IEEE.

Song I.,ITER Organization | Liu H.,ASIPP | Li J.,ASIPP | Gao G.,ASIPP | And 2 more authors.
IEEE Transactions on Applied Superconductivity | Year: 2014

ITER plasma requires the use of an internal vertical stabilization (VS) coil set as part of the system for active control of vertical position. The required peak current and voltage of VS power supply are of order of 60 kA and 2.4 kV. Also, it needs short voltage response time, less than 1 ms and highly transient power demand. This paper describes a conceptual design study for the in-vessel vertical stabilization coil power supply system to judge design feasibility and estimate its cost. Finally, the result will be used as design input data for the ITER Tokamak building and other interface systems. © 2013 IEEE.

Hyatt A.,General Atomics | Humphreys D.A.,General Atomics | Welander A.,General Atomics | Eidietis N.,General Atomics | And 9 more authors.
IEEE Transactions on Plasma Science | Year: 2014

The DIII-D plasma control system (PCS), initially deployed in the early 1990s, now controls nearly all aspects of the tokamak and plasma environment. Versions of this PCS, supported by General Atomics, are presently used to control several tokamaks around the world, including the superconducting tokamaks Experimental Advanced Superconducting Tokamak and Korean Superconducting Tokamak Advanced Research. The experimental challenges posed by the advanced tokamak mission of DIII-D and the variety of devices supported by the PCS have driven the development of a rich array of control algorithms, along with a powerful set of tools for algorithm design and testing. Broadly speaking, the PCS mission is to utilize all available sensors, measurements, and actuators to safely produce a plasma state trajectory leading to and then maintaining the desired experimental conditions. Often new physics understanding leads to new or modified control requirements that use existing actuators in new ways. We describe several important DIII-D PCS design and test tools that support implementation and optimization of algorithms. We describe selected algorithms and the ways they fit within the PCS architecture, which in turn allows great flexibility in designing, constructing, and using the algorithms to reliably produce a desired complex experimental environment. Control algorithms, PCS interfaces, and design and testing tools are described from the perspective of the physics operator (PO), who must operate the PCS to achieve experimental goals and maximize physics productivity of the tokamak. For example, from a POs (and experimental team leader's) standpoint, a PCS algorithm interface that offers maximum actuator, algorithmic, and measurement configuration flexibility is most likely to produce a successful experimental outcome. However, proper constraints that limit flexibility in use of the PCS can also help to maximize effectiveness. For example, device limits and safety must be built into the PCS, sometimes at the algorithm level. We show how the DIII-D PCS toolset enables rapid offline testing of a new or modified algorithm in a simulated tokamak environment. Finally, we illustrate usage of PCS-based checklists and procedures that enhance experimental productivity, and we describe an asynchronous condition detector system within the PCS that enhances device safety and enables complex experiment design. © 1973-2012 IEEE.

News Article | November 11, 2015

The team is led by Dr. Xianzu Gong of ASIPP and Dr. Andrea Garofalo of General Atomics (GA) in San Diego. Using both China's EAST facility and the DIII-D National Fusion Facility, operated by GA for the U.S. Department of Energy, the team has investigated the "high-bootstrap current" scenario, which enhances self-generated ("bootstrap") electrical current to find an optimal tokamak configuration for fusion energy production. Magnetic fusion energy research uses magnetic fields to confine plasma (ionized gas) heated to temperatures hotter than the Sun's core. This enables the ions to fuse and release excess energy that can be turned into electricity, harnessing the Sun's power on Earth. The most developed configuration is the tokamak, and the team's work helps prepare for the 500-megawatt ITER fusion research facility that is currently being built in France by a consortium of 35 nations, including China and the U.S. This joint U.S.-China experiment directly demonstrates the stabilizing effect of reducing the plasma-wall distance in tokamaks with high plasma pressure and large bootstrap current fraction, according to Dr. Gong, who said, "I think, in simple terms, these experiments may provide better physics and operation foundation for ITER plasmas." The focus was on resolving the "kink mode" instability, a wobbling effect that reduces performance, by moving the plasma closer to the vessel's wall, Dr. Garofalo explained . Operating closer to the wall suppresses the kink mode and enables higher pressure inside the tokamak, the toroidal or doughnut-shaped steel-lined fusion device. This gives rise to "pressure-driven" plasma flows that maintain the confinement quality even with lower external injection of velocity. "This is unlike any other regime," said Dr. Garofalo. "It's very risky to move the plasma that close to the wall. The chief operator said 'You can't do that anymore, you're going to damage the machine,' so it was a struggle to prove our theory was correct." The gambit paid off. Moving the plasma closer to the wall removed the kink mode and enabled higher plasma pressure, which, in turn, makes the plasma less dependent on externally injected flow. This is important because in a tokamak reactor, such as ITER, it is very difficult and expensive to drive a rapid plasma flow with external means. The team performed the most recent bootstrap exploration in DIII-D, following-up work on the record-setting milestone achieved at China's EAST tokamak, where GA scientists have also been collaborating. An ASIPP scientist Dr. Qilong Ren will deliver the invited talk on the topic of Magnetic Confinement-Experiments. While fusion has been in the public domain since the 1950s and its advances have been achieved by teams around the world, this U.S.-China team is setting new milestones in global cooperation. For realization of magnetic fusion energy, global cooperation is needed, said Dr. Gong of ASIPP, who cited the EAST/DIII-D partnership as "an efficient and effective new model" for international science collaborations that benefits both partners and the field of study. "We have made a very good start of international collaboration in fusion research between China and the U.S., and we are very proud to be a pioneer in this field," said Dr. Gong. More information: Abstract: KI2.00004 Progress Toward Steady State Tokamak Operation: Exploiting the high bootstrap current fraction regime Session Session KI2: MFE Regime Optimization

Liang L.Z.,ASIPP | Hu C.D.,ASIPP | Wei J.L.,ASIPP
IPAC 2013: Proceedings of the 4th International Particle Accelerator Conference | Year: 2013

Considering the beam divergence and the convergence of the spherical electrode, the beam transmission model is presented, and the variation of beam edge is described by a formula, which is used to calculate the beam divergence half-angle with the experimental data obtained by the thermocouples. Assuming the beam divergence half-angle is constant in space and time, the beam profile distribution formula and variation of beam axial intensity are introduced. Taking the HT-7 Diagnostic Neutral Beam (DNB) as a reference, the divergence half-angle is calculated for the neutral beam shot 60901. The 1/e half width of beam at collimation target calculated by formula is in agreement with that of experimental data. Variation of beam edge and axial intensity with downstream distance is estimated for HT-7 diagnostic neutral beam. Copyright © 2013 by JACoW.

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