Electrical insulation testing for ITER fusion tokamak

The uniqueness of the ITER fusion project drives purpose and scope here to fulfil (functional) Procurement Arrangement (PA) requirements for the Central Solenoid (CS) and Correction Coil (CC) magnets' electrical insulation [1-3]; as used and undertaken for covering high voltage testing operations of the magnet coil winding pack (WP) vacuum pressure insulation (VPI) beam qualification, followed in 2016 by start of magnets series production in the USA and China [1, 2]. Useful to be considered common state-of-the-art electrical power components testing practices [3, 4] are presented here as performed to verify that the integrity and insulation of the various components are within acceptable limits during different phases of the fabrication and to certify acceptance on delivery of the complete coils. Initial measurement plans were complemented into electrical tests with typical high voltage engineering implemented measures as a pre-requisite to successfully validate produced results. The above permitted the first quantitative validation of the obtained final production, including life time behavior

Testing of the ITER CS Module #4

The ITER Central Solenoid is under fabrication by the US ITER organization and its subcontractors. US ITER will supply seven modules to ITER IO, six of which will be assembled in a stack that forms the ITER Central Solenoid (CS). All CS modules were or will be tested at 40 kA in the Final Test facility at General Atomics, Poway, CA. Testing included high voltage, as well as Paschen testing in the vacuum and global leak tests before and after the cooldown to 4.5 K and EM cycling to 40 kA. In the paper we present the results of the CS Module 4 performance, after modifications to the test facility to improve reliability and instrumentation. We measured critical temperatures in several pancakes, AC losses before and after 10 cycles to 40 kA, joints resistance and hydraulic characteristics of the coils. We also measured displacements of the coil height and vertical strain of the CSM (central solenoid module) to verify structural mechanical characteristics of the coil along with cooldown shrinkage of the coil. We studied performance of the cowound quench detectors, confirmed their effectiveness in suppression of the inductive noise, but also developed a plan to improve sensitivity of the quench detection in ITER CS. This information is necessary for verification of the stack behavior of CS in ITER operation. The test results, preliminary analyses, comparisons to the other tested modules are presented and discussed

Testing of the ITER Central Solenoid Modules

The ITER Central Solenoid is under fabrication by the U.S. ITER organization and its subcontractors. U.S. ITER will supply seven modules to ITER IO, six of which will be assembled in a stack that forms the ITER Central Solenoid. The first modules that were built by GA at their facility, went into high voltage testing, including Paschen testing in the vacuum, and then they were tested at 4.5 K and up to 40 kA to demonstrate compliance of the coil with the ITER requirements. In this article, we present the Test Plan and results of the central solenoid (CS) module's performance, especially at the full current. We measured critical temperatures in several pancakes, we measured ac losses, joint resistance, and hydraulic characteristics of the coils. We also measured displacements of the coil height and hoop strain of the CS module (CSM) to verify the structural mechanical characteristics of the coil along with the cooldown shrinkage of the coil. We studied the performance of the cowound quench detectors and confirmed their effectiveness in the suppression of inductive noise. This information is necessary for verification of the stack behavior of CS in ITER operation. The test results and preliminary analyses are presented, compared to expectations, and discussed

First ITER CS module test results

The ITER Central Solenoid (CS) is under fabrication by the US ITER organization and its subcontractors. US ITER will supply seven modules to ITER IO, six of which will be assembled in a stack that forms the ITER Central Solenoid, with one as a spare. The first module fabrication has been completed by General Atomics (GA) at their facility and has begun testing including high voltage testing, Paschen testing in the vacuum and then testing at 4.5 K and up to 40 kA in order to demonstrate compliance of the coil to ITER requirements. In the paper we present the Test Plan and results of the CS Module performance tests, especially at 40 kA current. AC losses, joint resistances and hydraulic characteristics of the coil are all measured. Displacements of the coil height and hoop strain of the CS Module are also measured to verify structural and mechanical characteristics of the coil along with cooldown shrinkage of the coil. This information is used for verification of the stack behavior of CS in ITER operation. The test results and preliminary analyses results are presented, compared to expectations, and discussed. All measured parameters suggest that the CS module will perform well in ITER machine

DIII-D research advancing the physics basis for optimizing the tokamak approach to fusion energy

Funding Information: This material is based upon work supported by the US Department of Energy, Office of Science, Office of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Office of Science user facility, under Awards DE-FC02-04ER54698 and DE-AC52-07NA27344. Publisher Copyright: © 2022 IAEA, Vienna.DIII-D physics research addresses critical challenges for the operation of ITER and the next generation of fusion energy devices. This is done through a focus on innovations to provide solutions for high performance long pulse operation, coupled with fundamental plasma physics understanding and model validation, to drive scenario development by integrating high performance core and boundary plasmas. Substantial increases in off-axis current drive efficiency from an innovative top launch system for EC power, and in pressure broadening for Alfven eigenmode control from a co-/counter-I p steerable off-axis neutral beam, all improve the prospects for optimization of future long pulse/steady state high performance tokamak operation. Fundamental studies into the modes that drive the evolution of the pedestal pressure profile and electron vs ion heat flux validate predictive models of pedestal recovery after ELMs. Understanding the physics mechanisms of ELM control and density pumpout by 3D magnetic perturbation fields leads to confident predictions for ITER and future devices. Validated modeling of high-Z shattered pellet injection for disruption mitigation, runaway electron dissipation, and techniques for disruption prediction and avoidance including machine learning, give confidence in handling disruptivity for future devices. For the non-nuclear phase of ITER, two actuators are identified to lower the L-H threshold power in hydrogen plasmas. With this physics understanding and suite of capabilities, a high poloidal beta optimized-core scenario with an internal transport barrier that projects nearly to Q = 10 in ITER at ∼8 MA was coupled to a detached divertor, and a near super H-mode optimized-pedestal scenario with co-I p beam injection was coupled to a radiative divertor. The hybrid core scenario was achieved directly, without the need for anomalous current diffusion, using off-axis current drive actuators. Also, a controller to assess proximity to stability limits and regulate β N in the ITER baseline scenario, based on plasma response to probing 3D fields, was demonstrated. Finally, innovative tokamak operation using a negative triangularity shape showed many attractive features for future pilot plant operation.Peer reviewe