ITER Organization Research Articles

Numerical Simulation of ITER PF1 Coil Heating at VPI of Electrical Insulation

The ITER Organization imposes extremely high requirements for mechanical and dielectric strength of the electrical insulation of the superconductive coils because of a different nature of stresses in the ITER machine. The VPI method is required so as to make the insulation monolithic and durable. To provide a qualitative insulation, a temperature non-uniformity of 5 should be maintained throughout the electrical insulation during the VPI process. The PF1 coil is about 150 t in weight and consists of materials with different thermal conductivity. Besides, a complicated geometry of the coil adds complexity to VPI. Since the VPI is usually a long-term and irreversible process (within pot life of compound), the whole process should be controllable. The paper describes the developed calculation model of the ITER PF1 coil and presents the results of numerical simulation of this model at VPI of electrical insulation. The performed full-scale simulation makes it possible to reduce essentially the cost of qualification activities and a risk of obtaining a low-quality electrical insulation of large-scale items. The completed VPI of the ITER PF1 coil has verified the proposed method for coil heating and proved an acceptable agreement between the calculated and measured temperature distributions throughout the VPI procedure.

Effect of non-condensable gas in steam condensation at sub-atmospheric pressure condition

In the International Thermonuclear Experimental Reactor (ITER), a postulated Loss Of Coolant Accident (LOCA) in the Vacuum Vessel (VV) has to be managed with a pressure suppression system working at sub-atmospheric pressure. The operating conditions considerably differ from those experienced in the fission nuclear power plants such as BWR, since the ITER Tokamak works at very low pressure conditions and can withstand a maximum pressure of 0.15 MPa. For this reason, the pressure value must not exceed 10 kPa for a water temperature of 30 °C inside the Vapour Suppression Tanks (VSTs) that are the fundamental components of the Vacuum Vessel Pressure Suppression System (VVPSS). During a LOCA some non-condensable gases (mainly hydrogen and oxygen gases due to the water radiolysis or thermolysis) may be mixed in the steam and this could impair the condensation efficiency. In order to investigate the effects of non-condensable gas on DCC, we conducted a research program funded by ITER Organization at the laboratory of the University of Pisa: we designed and built a small-scale experimental rig to study the steam Direct Contact Condensation (DCC) with the presence of non-condensable gas and simulate the behaviour of a VST. Since DCC can occur with different characteristics, we ran 12 closed mode tests exploring all condensation regimes injecting a certain mass of air with the steam discharged in the subcooled water. The tests started at the saturation pressures corresponding to water temperatures ranging from 40 °C to 80 °C and ended when the free space volume reached the atmospheric pressure. From the analysis of the data acquired during the tests we observed that the condensation efficiency remained higher than 95%. We observed that despite this, the presence of a certain quantity of non-condensable gas has negative aspects on condensation: the condensation regime never reaches stability (the regimes quickly shift towards instability). Furthermore, the presence of air triggers a turbulence of the flow which interferes with the transfer of heat from the steam to the water. Not only the introduction of air into the flow increases linearly the pressure above the water head but its high temperature further contributes to the pressure increase. The air mixed with the flow also forms eddies that can trap the steam and transport it out of the water, preventing it from condensing. Apart from condensation, the most noticeable problem is the rapid increase in pressure inside the condensation tank.

Entraining of dust and steam flux in water pool: experimental and numerical simulations of vapor condensation and dust deposition

This paper deals with experimental tests of steam condensation in a water pool at atmospheric pressure and temperature in the range 15-100 °C. The activity is performed in the frame of a research program, funded by the ITER organization, for the study of dust deposition, produced in the ITER Vacuum Vessel and entrained by the steam and non condensable gas into the Pressure Suppression Tanks in the case of a Loss of Coolant Accident. The steam condensation into a subcooled water pool has been investigated to characterise the condensation regimes occurring during dust deposition tests. The dust distribution on the tank walls strongly depends on the steam jet length and on the effective heat transfer coefficient. Few grams of dust reduce the water transparency, therefore separated tests without dust overcome this drawback. Measurements of the lengths and surfaces of the steam jets (which permit to calculate effective average heat transfer) have been performed by means of image analyses and have been compared with theoretical correlations. The comparison showed a good agreement between the experimental data and theoretical correlations. Useful data have been obtained for implementing numerical models of dust deposition and for understanding the dust distribution on the tank wall obtained in the experimental tests.

"Everybody in this room can understand": A qualitative exploration of peer support during residency training.

Though peer support groups are often utilized during residency training, the dynamics, content, and impact of social support offered through peer support are poorly understood. We explored trainee perceptions of the benefits, drawbacks, and optimal membership and facilitation of peer support groups. After engaging in a peer support program at an emergency medicine residency program, 15 residents and 4 group facilitators participated in four focus groups in 2018. Interview questions explored the dynamics of group interactions, types of support offered, and psychological impacts of participation. The authors conducted a reflexive thematic analysis of data, performing iterative coding and organization of interview transcripts. Discussions with experienced senior residents and alumni normalized residents' workplace struggles and provided them with insights into the trajectory of their residency experiences. Vulnerable group dialogue was enhanced by the use of "insider" participants; however, residents acknowledged the potential contributions of mental health professionals. Though groups occasionally utilized maladaptive coping strategies and lacked actual solutions, they also enhanced residents' sense of belonging, willingness to share personal struggles, and ability to "reset" in the clinical environment. Participants offered insights into the benefits and drawbacks of peer support as well as optimal peer group composition and facilitation. Support groups may be more effective if they engage a complementary model of alumni and pre-briefed psychologist facilitators, avoid fatalism, and aim to foster intimate connections among residents. These findings can inform the development of future initiatives aiming to create a safe space for trainees to discuss workplace stressors.

Deposition of ITER Vacuum Vessel Dust Inside the Pressure Suppression System During a Loss of Coolant Accident: Experimental and Numerical Analyses

Abstract This paper deals with an experimental and numerical analysis of the deposition of International thermonuclear experimental reactor (ITER) dust simulant inside a reduced scale vacuum vessel pressure suppression system (VVPSS) of ITER. This research, funded by the ITER organization, aims to analyze the dust deposition in a water container in a relevant configuration for the ITER VVPSS, the dust removal by means of robotized apparatuses and their decontamination efficiency. The experimental rig, built at the department of civil and industrial engineering (DICI) of the University of Pisa (Italy), is described and the results of a preliminary experimental test are illustrated. It was measured that two and 82% of the overall injected dust mass was strongly and lightly bound to the water container surfaces, respectively, and 16% were not deposited in the water tank. Furthermore, three numerical analyses were carried out implementing a model of the experimental rig in the Enel code for analysis of radionuclide transport (ECART) to determine the relevance of different parameters on the deposition, resuspension and removal of dust. The numerical simulations allowed to specify dust mass deposition on the different rig components, revealing a strong dust retention (about 66%) in the first part of the injection piping in case of coarse granulometry. Finest lognormal dust distribution was instead able to reach the water container (about 90%). Moreover, the numerical simulations permitted to define more precisely the test matrix and to analyze the experimental results.

The upgrade of data processing and storage system for EAST NBI

Inspired by the ITER organization's CODAC system, the current experimental system architecture is being progressively redesigned using the experimental physics and industrial control system (EPICS) to meet the needs of future long pulse experiments. As part of the overall experimental system, the data processing and storage system (DPSS) is mainly used to provide services of data processing and data storage for the experimental system. However, with the data processing mechanism of the previous DPSS, the previous experimental system is unsatisfactory in terms of real-time experimental data feedback, and the data processing mechanism of the previous DPSS is not compatible with the new architecture. Given that the original DPSS no longer fits the current experimental characteristics in terms of both functionality and demand so that it is necessary to redesign and upgrade the original DPSS to adapt to the new architecture. In view of the soft real-time and distributed characteristics of EPICS, the upgraded DPSS will be redesigned based on the data processing mechanism of EPICS. And it uses an adaptive strategy that allows the experimental system to automatically match more appropriate data uploading modes according to the rate of data acquisition. For this purpose, a data transmission protocol based on the TCP/IP protocol has been proposed. With regard to software optimization, the upgraded DPSS uses technologies of double-buffer and MMAP to optimize data exchanging between memory spaces. In terms of data storage, the upgraded DPSS adopts the segment strategy to manage experimental data efficiently. In addition, to prevent problems caused by abnormal circumstances, a guarantee mechanism for data storage is provided in the upgraded DPSS. After being tested, the upgraded DPSS will later be integrated into the general control system of EAST NBI.

Computational design of magnetic beamlet deflection correction for NNBI

Coextracted electrons are inevitable for negative ion based NBI systems. To avoid wastefully accelerating these electrons to the full energy, they are magnetically deflected out of the beamlet onto the extraction grid. The deflection of the electrons creates an unwanted deflection of the ions, which is magnetically corrected in the ITER heating neutral beam. In order to gain experience with magnetic deflection correction, IPP has designed a magnetic deflection correction system in collaboration with the ITER organization, which will be characterized in the BATMAN Upgrade test facility. The ion-optics code IBSimu is used for the design of the magnetic deflection correction system. A comparison of the design candidates shows that the magnetic deflection correction system inherently leads to an increased divergence as result of the magnetic field structure. The divergence increase manifests itself as an increased fraction of a broader halo component.

Entanglement devised barren plateau mitigation

Hybrid quantum-classical variational algorithms are one of the most propitious implementations of quantum computing on near-term devices, offering classical machine-learning support to quantum scale solution spaces. However, numerous studies have demonstrated that the rate at which this space grows in qubit number could preclude learning in deep quantum circuits, a phenomenon known as barren plateaus. In this work, we implicate random entanglement, i.e., entanglement that is formed due to state evolution with random unitaries, as a source of barren plateaus and characterize them in terms of many-body entanglement dynamics, detailing their formation as a function of system size, circuit depth, and circuit connectivity. Using this comprehension of entanglement, we propose and demonstrate a number of barren plateau ameliorating techniques, including initial partitioning of cost function and non-cost function registers, meta-learning of low-entanglement circuit initializations, selective inter-register interaction, entanglement regularization, the addition of Langevin noise, and rotation into preferred cost function eigenbases. We find that entanglement limiting, both automatic and engineered, is a hallmark of high-accuracy training and emphasize that, because learning is an iterative organization process whereas barren plateaus are a consequence of randomization, they are not necessarily unavoidable or inescapable. Our work forms both a theoretical characterization and a practical toolbox; first defining barren plateaus in terms of random entanglement and then employing this expertise to strategically combat them.

The JOREK non-linear extended MHD code and applications to large-scale instabilities and their control in magnetically confined fusion plasmas

JOREK is a massively parallel fully implicit non-linear extended magneto-hydrodynamic (MHD) code for realistic tokamak X-point plasmas. It has become a widely used versatile simulation code for studying large-scale plasma instabilities and their control and is continuously developed in an international community with strong involvements in the European fusion research programme and ITER organization. This article gives a comprehensive overview of the physics models implemented, numerical methods applied for solving the equations and physics studies performed with the code. A dedicated section highlights some of the verification work done for the code. A hierarchy of different physics models is available including a free boundary and resistive wall extension and hybrid kinetic-fluid models. The code allows for flux-surface aligned iso-parametric finite element grids in single and double X-point plasmas which can be extended to the true physical walls and uses a robust fully implicit time stepping. Particular focus is laid on plasma edge and scrape-off layer (SOL) physics as well as disruption related phenomena. Among the key results obtained with JOREK regarding plasma edge and SOL, are deep insights into the dynamics of edge localized modes (ELMs), ELM cycles, and ELM control by resonant magnetic perturbations, pellet injection, as well as by vertical magnetic kicks. Also ELM free regimes, detachment physics, the generation and transport of impurities during an ELM, and electrostatic turbulence in the pedestal region are investigated. Regarding disruptions, the focus is on the dynamics of the thermal quench (TQ) and current quench triggered by massive gas injection and shattered pellet injection, runaway electron (RE) dynamics as well as the RE interaction with MHD modes, and vertical displacement events. Also the seeding and suppression of tearing modes (TMs), the dynamics of naturally occurring TQs triggered by locked modes, and radiative collapses are being studied.

Large scale experimental facility for performance assessment of the vacuum vessel pressure suppression system of ITER

The nuclear fusion reactor ITER (International Thermonuclear Experimental Reactor) foresees a Pressure Suppression System (PSS) in order to manage a Loss of Coolant Accident (LOCA) in the Vacuum Vessel (VV). A LOCA scenario is postulated to occur in the cooling systems of plasma facing components. The Vacuum Vessel Pressure Suppression System (VVPSS) has a key safety function because a large internal pressure in the VV could lead to a breach of the primary confinement barrier.Steam condensation would occur in ITER at sub-atmospheric pressure, differently than the applications in nuclear fission reactors. Steam direct contact condensation (DCC) in sub-atmospheric conditions have never been implemented before, therefore an experimental assessment of the VVPSS is necessary to simulate this phenomenon in ITER's thermal-hydraulic conditions.Under the financial support of ITER Organization, an experimental program was performed at the Department of Civil and Industrial Engineering (DICI) of the University of Pisa in the past 3 years, using a small-scale experimental rig (1/22 scale factor).In order to assess the scale laws elaborated by means of the experimental results obtained in this small scale facility, a Large Scale facility was built at University of Pisa. This facility is almost a full scale mock-up of ITER's VVPSS from the geometric point of view (1/1.09 scale factor) and it is 1/10 scale factor from the steam mass flow rate point of view.This paper illustrates the results of the first test campaign. A 1/10 scale sparger system SPS-A (having a diameter equal to DN125 and 100 holes) was used with a maximum steam mass flow rate of 0.5 kg/s. This experimental configuration represents real operating conditions on the basis of the results of elaborated similitude analysis. The preliminary analysis of the experimental results confirms the agreement in terms of condensation regimes between the two reduced scale (1/22 and 1/10 scale) as foreseen by the elaborated scale laws.

Development of Coax Compacted Joint Assembly Process for the ITER Central Solenoid

The Central Solenoid (CS) is the core element of the ITER magnet system, contributed as in-kind procurement by the US Domestic Agency. Made up of six modular inter-exchangeable module coils, vertically stacked, forming a 15 m high and 4 m outer diameter solenoid, to be inserted inside the central tokamak core, after assembly of the 18 D-shaped Toroidal Field (TF) coils. Each module uses a Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn conductor internally cooled by circulation of supercritical helium at 4.5 K and supplied, for each module, with a 45 kA current by the way of two vertical leads exiting from the module outer radius. The available space between the CS and TF magnets being very limited, the US DA has developed a compact - so called Coax Joint (CJ) - devoted to connect the respective twelve module leads to the feeders located at top and bottom of the CS via dedicated busbar extensions. The CS coax joint assembly procedure as developed by US DA would make use of CS assembly onsite soldering process, to be executed under supervision of the ITER Organization (IO) in the tokamak assembly hall. In order to mitigate risks related to these soldering activities needed at assembly stage, IO has proposed an alternative assembly process based on indium wires compaction - so called Coax Compacted Joint (CCJ)- and initiated its prototype development and qualification in time. The hereby presented CCJ design solution is based on four copper quadrants, each including an embedded straight Rutherford-type superconductor cable-strip to transport the current. The current is transferred from the lead to the quadrants by the way of compacted indium wires. A steel jacket welded around the joint ensures the mechanical support as well as the needed leak tightness. The paper describes developments made by CEA under IO supervision from the conceptual design to the testing of joints in relevant cryogenic and operative conditions.

Strategy for D-shape assembly of ITER vacuum vessel sector #06 as applying 3D metrology

The first sector of the ITER vacuum vessel (VV) has been delivered to the ITER Organization. One of the critical manufacturing steps is d-shape assembly. The manufactured four poloidal segments (PS 1, 2, 3, 4) are assembled, and then form the d-shape structure with a height of 11.4 m. Each PS is a complex and heavy double-wall structure with an inner shell and outer shell, poloidal and toroidal ribs, and in-wall shielding blocks are filled inside of the double wall as radiation shield. Thus, it is quite difficult to handle for assembly in keeping with the segment final dimensional status, so that it is necessary to set a practical plan. Before conducting actual assembly, the d-shape is virtually fitted using all results of final 3D metrology on each PS. In detail, each inner shell as-built result on four interfaces, plus a margin for welding shrinkage and gap for welding are considered to set the d-shape fit-up reference. To set each PS in the right position as comparing the fit-up reference, the fiducial posts (FPs) on each PS are measured using a 3D metrology system. After done the fit-up on the assembly plate, whole FPs (55 ea.) are checked as verification. As a result, the PS2 and PS4 are shifted Avg 7.4 mm and Avg 7 mm to the poloidal way, respectively. After finished inner shell welding, the whole FPs are also measured. The PS2 and PS4 are still shifted Avg 1.4 mm and Avg 0.8 mm to poloidal way, respectively. Eventually, the as-built dimension on the important part is evaluated with given tolerance as a preliminary prediction of the final d-shape result.

AC Losses in the First ITER CS Module Tests: Experimental Results and Comparison to Analytical Models

The ITER Central Solenoid (CS) will be manufactured by assembling a stack of six modules, which are under fabrication by the US ITER organization and its subcontractors. The tests of the first CS Module have been performed at the premises of the General Atomics (GA) facility in Poway (US), in order to check compliance to the ITER requirements. Among other tests, the magnet was submitted to exponential dumps of the transport current from different initial values (10, 15, 20, 22.5, 25, 35, 40 kA) down to 0 kA. These tests are aimed at conducting DC breaker commissioning of the test facility and were used to measure the AC losses in the coil during electrodynamic transients. This paper presents the results of these measurements, along with a comparison with analytical computations of the losses in the magnet.

Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System

The Water Cooled Lithium Lead Test Blanket System (WCLL TBS) is one of the EU Test Blanket Systems candidate for being installed and operated in ITER. In view of its Conceptual Design Review by F4E and ITER Organization (IO), planned for mid-September 2020, several technical activities have been performed in the areas of WCLL TBS Ancillary Systems design.In this article the outcomes of the conceptual design phase of the four main Ancillary Systems of WCLL TBS, namely the Water Cooling System (WCS), the Coolant Purification System (CPS), the PbLi loop and the Tritium Extraction System (TES), are reported and critically discussed.In particular, for each Ancillary System hereafter are reported: i) a short design description, including the conceptual design of their main components together with their operative conditions under the so-called Normal Operational State (NOS), ii) the ESP-ESPN classification for their main components, and iii) their arrangement and integration in the assigned ITER areas (PC#16, Vertical Shaft, TCWS Vault, Galleries and Tritium Process Room).

Thermal Hydraulic Behavior of the First ITER CS Module

The ITER Central Solenoid (CS) modules are under fabrication by the US ITER organization and its subcontractors. US ITER will supply seven modules to the ITER Organization (IO), six of which will be assembled in a stack that forms the ITER Central Solenoid, the last one being spare. The first module, namely CSM 1 was manufactured by General Atomics (GA) and went through Factory Acceptance Tests (FAT) including high voltage testing, Paschen testing and then cold test at 4.5 K and up to 40.0 kA in order to demonstrate compliance with coil performance requirements. The paper focuses on the results of the first CS Module thermal hydraulic characterization without current (and field). Pressure drops, transit time, and thermal coupling between pancakes are presented. Analysis results and tests are also compared.

Circuit Controller for PF Converter System Based on CODAC Core System

The poloidal field converter system of the International Thermonuclear Experimental reactor, which is developed, assembled, and tested by the Institute of Plasma Physics of the Chinese Academy of Sciences, uses the CODAC core system provided by ITER organization to design its control system. This paper completes the design of the circuit control system as well as realizes the transmission, state transition, and real-time control of the operating status of the poloidal field converter system to meet specific requirements for the control cycle of the control system. This paper provides solutions to some key problems, such as the state transition, real-time transmission, and operating state display, and real-time control of the converter circuit control system. The system designed by this paper has been successfully tested in the ITER PF prototype and ITER PF test platform.

Model-based real-time surface heat flux and temperature estimation for the DIII-D tokamak

A control-oriented model for monitoring of wall power flux densities on the DIII-D tokamak has been successfully implemented and validated experimentally. Future reactors will have to withstand severe steady state high heat flux loads on plasma-facing components (PFCs). Due to the difficulty of directly-measuring local heat fluxes on these components, monitoring and protection of PFCs during the plasma discharge can benefit from simplified physics-based real-time functional models to estimate and guide heat load control. As a first step into the development, a control-oriented model for monitoring of wall power flux densities and temperatures on DIII-D tokamak has been successfully implemented. The paper discusses the experimental demonstration and comparison of the 2D model-based wall heat flux algorithm on the DIII-D inner wall limiter (IWL) against infra-red (IR) camera heat flux measurements for limited plasma configurations. The paper also reports on the benchmarking of the field line tracing environment, SMITER, developed at ITER organization on DIII-D tokamak against experimental IR diagnostic data and the derivation of the component shaping weighting factors for the 2D model-based approach. Extension of the model-based approach for surface temperature estimation on the DIII-D IWL is also presented.

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.

Experimental results of multiple shattered pellet injection systems in KSTAR

Shattered pellet injection (SPI) is the technology chosen for the ITER Disruption Mitigation System and is explored at several fusion research devices, like DIII-D and JET and J-TEXT. The ITER disruption mitigation strategy relies on multiple injections to achieve RE (runaway electron) avoidance with optimum TQ (thermal quench), CQ (current quench) durations to adequately reduce wall loads. To demonstrate the feasibility of the multiple injection and to extrapolate to ITER, experiments with two identical injectors toroidally opposite to each other are needed urgently. KSTAR (Korea Superconducting Tokamak Advanced Research) can be a unique testbed to study the plasma disruption mitigation for ITER. KSTAR has installed two identical injectors in 180 degrees of toroidal opposite positions in 2019. For this system, ORNL (Oak Ridge National Laboratory) provided the two injectors, the shatter tubes, and auxiliary systems. NFRI (National Fusion Research Institute) provided the infrastructure of a vacuum pumping system, control & data acquisition system, and installed additional diagnostic systems for SPI in collaboration with the ITER Organization. This paper describes the engineering achievements during installation on KSTAR and the initial results of single and multiple SPI experiments in the 2019 campaign.

Progress of water cooled ceramic breeder test blanket module system

A Water-Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) System is being developed as one of the most important steps toward DEMO blanket in Japan. As for the ITER-TBM program, the conceptual design of WCCB-TBM system has been approved by the ITER organization (IO). And two years have already passed after the start of the preliminary design activity. WCCB TBM Team had a concern about TBM box structure withstanding over 15 MPa of the coolant pressure, and decided to change the design as the result of study on structural integrity. Recently, the design change of the TBM structure has been approved domestically. So, WCCB TBM Team has applied the design change to the IO, and agreed with the IO to receive the conceptual design review which is limited the scope to the TBM structure. This paper provides an overview of the recent achievements of the development of the WCCB Test Blanket Module System in Japan.