Central Solenoid Modules Research Articles

Structural Qualification of ITER Central Solenoid Module During Transport From California, USA to ITER Site

Structural Qualification of ITER Central Solenoid Module During Transport From California, USA to ITER Site

Exploration of ITER operational space with as-built stiffness of central solenoid modules

The as-built stiffness in the ITER central solenoid (CS) modules (CSM1 thorough to CSM4 are currently manufactured) determines the range of vertical compression forces that can be tolerated by the CS modules during ITER operation. Since the as-built stiffness of the CS modules manufactured (∼32 GPa and ∼34 GPa for CSM1 and CSM2, respectively and similar for the other modules) has been reduced from the design value (53 GPa), the CS axial (vertical) force criteria have been updated assuming a conservative stiffness (25 GPa) with margins for all six CS modules. Initial analysis using the updated CS force criteria has revealed that this reduction affects only the plasma initiation with fully charged CS in the ITER 15 MA Baseline DT scenario, resulting in a slight reduction of poloidal magnetic flux, from 117.5 Wb to 116.2 Wb at initial CS magnetization. Therefore, the 15 MA Baseline scenario has been re-developed with an updated plasma start-up, and then the entire evolution of the CS and poloidal field coil parameters has been validated against all the coil currents, fields and forces criteria. To explore potential risks and opportunities for further optimization of scenarios, the equilibrium operational space (the plasma internal inductance versus the poloidal magnetic flux produced by the coils) at flat-top burn has been analyzed using the CORSICA and DINA codes. The three major ITER reference DT operation scenarios, 15 MA Q = 10 Baseline, 12.5 MA Q > 5 Hybrid and 10 MA Q ∼ 5 Steady-State, satisfy all the coil criteria including the CS force updated reflecting the as-built stiffness. The evolution of the plasma discharge parameters within the equilibrium operational spaces provided a guidance for potential optimization with margins.

Estimation of Cool-Down Speed Affected by Mass Flow Rate and Inlet Pressure in JT-60SA CS Module

The JT-60SA central solenoid (CS) module is cooled to 4.5 K using liquid helium. This cool-down operation of the coil is planned to take about one month. During cool-down, the maximum temperature difference in the coil (dT <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">C-M</sub> ) must be kept below 40 K to avoid excessive thermal stress. Thermometers are attached only at the inlet and outlet on the outer perimeter of the coil. However, because the maximum temperature inside the coil occurs near its inner edge, dT <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">C-M</sub> cannot be measured directly. Hence, estimating dT <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">C-M</sub> is important to allow the cool-down process to be completed within one month without excessive thermal stress on the CS. In the present study, based on the results of the JT-60SA cool-down test performed in 2020, a temperature simulation model of the CS module, including the strands, jacket, and He in the bundle and hole regions was created using MATLAB, and dT <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">C-M</sub> was estimated during the cool-down test. This model was then used to evaluate the effect of the mass flow rate and inlet pressure on the cool-down speed.

Validation of the 4C Code on the AC Loss Tests of a Full-Scale ITER Coil

The AC loss tests on the first ITER Central Solenoid Module (CSM) have been modelled and compared to the test results. The model has been implemented in the 4 C code, a thermal-hydraulic modelling tool which includes the CSM winding pack and the cryogenic circuit of the test facility. Two modes of operation of the circuit have been analyzed: the nominal and the “isolation” mode, i.e., when the cryogenic circuit valves are operated to isolate the coil during the current dumps. The computed mass flow rate, pressure and coil outlet temperature at different locations have been compared with the measurements, showing a very good agreement in both modes of operation of the circuit. The validated model helped in the interpretation of the experimental results, such as the backflow at the coil inlet -which cannot be measured- or the non-monotonic outlet temperature evolution following the current dump. Furthermore, the code was used to qualify the isochoric method for the quantification of the deposited energy due to AC losses, as it was the only method applicable in case of current dumps from high current.

Analytical Modeling of Magnetization Losses in Twisted Stacked HTS Conductors

The European post-ITER fusion demonstrator reactor (DEMO) design activities have been ongoing for several years in the framework of the EUROfusion Consortium. The designs proposed for the DEMOnstration Fusion Power Plants (DEMO) magnets include options based on high-temperature superconducting (HTS) technology, namely as an Insert in the Central Solenoid (CS). This work describes the development of tools for the calculation of magnetisation losses in a twisted-stacked HTS cable, which represents one of the most promising configurations of HTS fusion Cable-in-Conduit conductors (CICC). This geometry cannot be tackled with the straightforward application of 1D or 2D models, given their inability to describe the twisting of the tape stacks. However, the methodology proposed in this work allows one to treat the twisting of the conductors by discretizing them in a set of straight pieces. A set of analytical equations is developed to compute the instantaneous power losses in the slab approximation. This fast, easy-to-use, open-access tool, can be employed for the computation of AC losses in the central solenoid modules of a tokamak. In particular, we have applied the model to study the HTS insert of the DEMO central solenoid.

Design of superconducting tokamak magnet systems and applications to HTS

In this paper we illustrate a self-consistent procedure to design the superconducting Toroidal Field (TF) and Central Solenoid (CS) magnet systems starting from the basic plasma dimensioning parameters. By defining the plasma major radius and minor radii, as well as the inboard TF coil neutron shielding radial built, we proceed by sizing the TF Winding Pack (WP), its case and the CS modules. The ITER magnets design philosophy and criteria are adopted throughout the study. A key outcome of this work is to illustrate the inter-play between machine size, magnetic field and availability of inductive flux for burn. In doing so we investigate the advantages brought by high field tokamak design compared the more conventional (ITER-like) Nb3Sn based magnet design solutions.

AC Loss Modeling of a Full-Size ITER CS Module

Thorough analyses of electromagnetic transients in superconducting coils for fusion magnets are necessary to compute the losses arising in time-varying operating conditions and correctly design the cryogenic system. The challenge of these analyses, when using numerical methods, is related to the high number of degrees of freedom required to discretize large-scale magnets wound with cable in conduit conductors (CICCs), which include hundreds of composite strands. This study is focused on the central solenoid (CS) of the ITER magnet system, composed of six modules, each wound with several CICC for a total length of 6 km. The numerical code THELMA and an analytical model based on a <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">single time constant</i> —previously applied to analyze the tests of a 44-m-long single-layer solenoid, referred to as CS Insert (CSI)—are scaled up to calculate losses during factory acceptance tests (FAT) at the General Atomic (GA) in Poway (US). We demonstrate, by comparison with experimental data, the possibility of predicting the losses in the CS module by projecting the results obtained in the CS Insert subsize test, a single-layer solenoid tested at the National Institute for Quantum Science and Technology (Naka, Japan). The application of a numerical model discretizing the conductor with a few tens of subcables to a full-scale fusion magnet is very demanding. A novel methodology is proposed to abate the computational burden, based on reducing the number of turns involved in the calculation, still retaining a high level of accuracy in the loss assessment.

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.

Development and validation of the 4C thermal–hydraulic model of the ITER Central Solenoid modules

The ITER Central Solenoid (CS) consists of a stack of six modules, each made of 40 pancakes wound with Nb3Sn Cable-In-Conduit Conductors (CICCs) cooled with supercritical helium (SHe). All six modules (plus one spare) are to be individually cold-tested at the General Atomics final test facility in San Diego (USA), in order to check their performance; the first CS Module (CSM1) was tested in early 2020.A test campaign on a CSM Mock-up (CSM MU) wound with 16 dummy pancakes, i.e., with non-superconducting (copper) strands, was already carried out in San Diego at the end of 2017, for the commissioning of the test facility. The analysis of the CSMMU experimental data is presented here.Each CSM is a full magnet with 554 turns; it did not have any thermal–hydraulic (TH) or electrical sensors inside the winding due to insulation reasons, so that, e.g., SHe pressure, temperature and mass flow rate, as well as the voltage, were only measured at the ends of selected pancakes.Therefore, it was essential to employ a thermal–hydraulic (TH) model in order to obtain information on the quantities of interest inside the coil, e.g. which was the voltage across the coil at the moment when the current sharing temperature (TCS) was reached for the first time somewhere in that double-pancake (DP) during a TCS test.The TH model of the CSM, developed and implemented in the validated 4C code, and eventually adopted for the test preparation and interpretation, includes some free parameters, i.e., the inter-pancake and inter-turn thermal coupling, whose uncertainty is mainly due to the complex, multi-layer structure of the turn and pancake insulation. The calibration of these parameters is required to correctly capture the TH behavior of the CSM. For this purpose, the results of the experimental campaign on the CSM MU have been used. The detailed topology of the CSM MU is described and implemented here in a dedicated 4C model. Both slow and fast transients are used for the calibration, e.g., quasi-steady state heating of the SHe, entering a single DP and heat slug tests, respectively. It is shown that the transverse heat transfer within the winding pack could be largely overestimated if the ideal heat conduction across a bulk insulation layer is considered. The calibrated model is then validated on the CSM1 test results.

Investigation of Transient Response Caused by Power Supply on JT-60SA Central Solenoid

The JT-60SA power supply of the central solenoid (CS) module contains multiple frequency components, and the transient response phenomenon and the resonance phenomenon cause the voltage non-uniformity in the CS. There is thus a possibility that a local voltage between the conductors that is larger than the withstand voltage of the insulation between conductors will be occurred, and the insulation is damaged. Therefore, an estimation of the transient response phenomenon and the resonance phenomenon is important for verifying that the CS is electrically stable. In the present study, the power supply voltage of the JT-60SA CS module was measured, and the effect of the transient response and the resonance phenomenon on the voltage between CS conductors was evaluated using an analysis model that included CS (four modules), the current lead between the power supply and the CS, and the structures surrounding the CS. The simulation results show that when the actual power supply is applied to the CS, the maximum voltage between CS conductors affected by the transient response and the resonance phenomenon is lower than the withstand voltage of the insulation between CS conductors. The results of our analyses provide fundamental data that can be used to estimate the voltage distribution in the CS and thus ensure the electrical stability.

AC Losses in the Second Module of the ITER Central Solenoid

The ITER Central Solenoid (CS) will be realized by assembling a stack of six modules. Each module is a solenoid consisting of 40 pancakes wound with a Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn Cable in Conduit Conductor (CICC). The tests of the second module (CSM#2) were performed at the General Atomics (GA) facility in Poway (US). During the test campaign, the CS Module is submitted to dumps of the transport current from different initial values (10, 15, 20, 25, 30, 35, 40 kA) to 0 kA, which allow measuring the AC losses in the coil. In this paper we present the results on AC losses during the dumps from different initial transport currents, as computed with two different methods. The first method is based on the observation that the dumps determine a very fast pressure rise of the supercritical helium embedded in the module, which undergoes an isochoric transformation. This approach is therefore based on the computation of the variation of internal energy of the helium during the pressure rise itself. The second method is based instead on a calorimetric procedure aimed at estimating the enthalpy variation of the supercritical helium due to the thermal power deposited during the current dumps. The main contribution to the losses during these dumps is due to the coupling losses. As for the hysteresis losses in the magnet, the results obtained through slow current cycle tests are also presented.

Examination and Characterization of Physical and Mechanical Properties of the ITER Central Solenoid Module Coils

The ITER Central Solenoid (CS) consists of a stack of six independent coil packs called modules. It features a total height of 18 m and a diameter of over 4 m. The modules are in an advanced stage of fabrication and testing by the US ITER Project Office (USIPO) and its subcontractor General Atomics (GA). A qualification module mockup at one to one scale but of reduced height was wound and Vacuum Pressure Impregnated (VPI) by GA to validate final manufacturing, using tooling and processes fully representative of a series module. The module was submitted to a thermal cycle down to the temperature of 4.5 K at which the coils will be cooled by supercritical helium. During plasma operation, the CS modules are subjected to a complex combination of static and dynamic forces. The understanding of the mechanical behaviour of the CS module coils is of paramount importance to analyse and predict the overall response of the CS stack. To this purpose, an extensive programme of investigation of the module mockup has been defined and applied. This allowed assessing, through examination and testing of a large number of VPI conductor array samples extracted from the mockup, the soundness of the coil through advanced non-destructive examination techniques including X-ray microtomography, dimensional metrology measurements and micro-optical observations. Moreover, additional testing of physical and mechanical properties carried out at room and cryogenic temperature allowed the behaviour of the conductor stacks to be assessed. The paper summarises the results of these investigations and their interpretation through mechanical analyses based on the individual properties of the coil constituents.

Start of the ITER Central Solenoid Assembly

The Central Solenoid (CS) is a key element of the ITER Magnet system, including six identical coils, called modules, assembled together to form a 4 m outer diameter, 13 m high solenoid. It is a superconducting magnet, using a 45 kA 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 with a peak field up to 13 T. It is enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components and the special assembly tooling of the ITER CS is the responsibility of US ITER, the ITER Domestic Agency of the USA, while the ITER Organization (IO) will carry out the assembly of these components. US ITER has awarded several contracts since 2011 to supply seven modules, including a spare, structure components, and the special tooling required for the CS pre-assembly. All deliveries are organized with the objective to start the CS assembly at IO site by the end 2021. IO has organized special process training and is now starting first phases of the assembly mostly focused on on-site assembly contractor. The paper describes the CS module manufacturing and delivery status, reports the special process final development and training, and assembly status. In particular, the outcomes from first modules factory acceptance tests results will be reported as well as the development and training for special activities to start the first module stacking. The bus bar joint and pre-compression processes with their related tooling qualifications will be detailed.

An analytical model for coupling losses in large conductors for magnetic fusion

During the operation scenarios of large superconducting tokamaks such as JT-60SA or ITER, disruptions or VDE (Vertical Displacement Events), AC losses are deposited in the superconducting magnets. Due to these AC losses, the temperatures of the Cable in Conduit Conductors (CICC) increase, reducing the temperature margins. It is of highest importance for the operation to correctly estimate AC losses and conductor temperatures. Due to the fast transient nature of those events, an important component of the losses is the coupling losses.The approach for coupling losses estimation, using a unique effective time constant “nτ” is not completely adequate to calculate these losses.For this purpose, since 2010, a heuristic analytical model called MPAS (Multizone PArtial Shielding), has been proposed. It is now currently used by several teams.For each conductor of the coils of a tokamak, the parameters of the MPAS model and nτ are estimated based on the characterization of the conductor AC losses under a sinusoidal excitation at different frequencies. This process is extensively conducted in existing facilities such as the SULTAN test facility at SPC (Swiss Plasma Center), on the dipole and press of Twente University or on JOSEFA dipole test facility at IRFM (Research Institute for Magnetic Confinement) at Cadarache.This characterisation leads to the constitution of a database for the different conductors used in the tokamak. This has been pushed to the uttermost within the ITER qualification and QA/QC (Quality Assurance, Quality Control).The methodology for estimating the MPAS parameters in relation with the measured AC losses is described in details and illustrated for two conductors. Using this database, it is then possible to calculate the coupling losses during ITER module discharges and scenarios. Estimation of coupling losses are compared to experimental results in recent experiments on ITER CS (Central Solenoid) modules.

Effect of SHe Temperature on Cool-Down Speed in JT-60SA CS Module

The JT-60SA central solenoid (CS), which consists of four stacked modules, is cooled to an operating temperature of 4.5 K by supercritical helium (SHe). During cool-down from room temperature to 4.5 K, in order to avoid a damage on the coil, it is required to reduce thermal stress due to temperature difference in the coil. In the cool-down operation, the CS needs to control the maximum temperature difference in the longitudinal direction of the cable-in-conduit (CIC) conductor below 40 K. However, since that could damage the coil, correctly estimating the temperature distribution in the conductor is important for ensuring the cool-down operation is performed safely. In the present study, the JT-60SA CS module was prepared and the inlet and outlet temperatures, as well as the SHe mass flow rate, were measured until the CS module was cooled to operating temperature, after which a CS module simulation was performed in order to determine the effect of the inlet temperature on the cool-down speed. From the simulation results, it was concluded that the average cool-down speed was 0.73 K/h and the CS module reached 80 K in approximately 8.3 days while maintaining the maximum temperature difference within 40 K. Furthermore, the temperature difference between the inlet and outlet was kept within 25 K in order to limit the thermal stress. Taken together, the results of our analyses provide fundamental data that can be used to evaluate the safe cool-down operation of the coil and thus protect the coil systems.

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.

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.

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.

ITER assembly phase: Progress toward first plasma

By the end of 2019, ITER had completed about 67% of the overall work required to achieve First Plasma. Significant progress has been achieved since SOFT 2018, most visible in the completion of key buildings and infrastructure on the worksite. The tokamak building will be ready for equipment as of Spring 2020. Commissioning of the connection to the EU grid and the substation for the steady-state electric network is complete. Planning for commissioning of other key plant systems (e.g., secondary cooling water, cryoplant, pulsed power for magnet systems) is well underway. On the manufacturing front, progress is equally impressive. The base and lower cylinder of the cryostat were completed and turned over to ITER in July 2019. The first two poloidal field coils will be turned over to ITER in early 2020, and two more are being wound. Factory tests of the first central solenoid module will be completed in 2020, and five more modules are in the manufacturing process. The first toroidal field coil is to arrive onsite in early 2020, and progress is rapid on additional TF coils. The first sections of the thermal shield are already onsite. The first vacuum vessel sector is complete and will arrive onsite in early 2020. This progress means that, for most of the major first-of-a-kind components, the capability has now been demonstrated to fabricate according to ITER’s demanding specifications. With this progress achieved, ITER will enter Assembly Phase in 2020, remaining on schedule for First Plasma in 2025. The sequence of ITER operation from First Plasma (FP) to the achievement of the Q = 10 project goal has been consolidated in a Staged Approach, with all systems to be installed before the start of full fusion power operation in 2035. The ITER Research Plan, issued publicly in 2018, is consistent with the systems available in each phase. An adjustment of the machine baseline configuration was made in 2019, with two equatorial ports allocated to the Test Blanket Systems and the development of a better performing Disruption Mitigation System. Highlights from each of these areas (manufacturing, commissioning, tokamak assembly) will be presented along with the updated status and plans.

Mechanical Pre-Compression System of the JT-60SA Central Solenoid

The mechanical loading technology has been developed for the pre-compression of central solenoid for the tokamak device of JT-60 super advanced (JT-60SA) being constructed in Japan. Four central solenoid modules were manufactured one by one, then 4 modules were stacked. The pre-compression were conducted to bind stacked 4 modules by support structure called tie-plate as a final process of the central solenoid manufacturing. The pre-compression jigs were designed by using the finite element analysis taking into account of plastic deformation of the jigs and the tie-plate. The pre-compression was performed using developed jigs. The hydraulic jacks and pumps were separately installed to the each tie-plate to control the pre-compress load individually. Since the pre-compress load changed by insulation creeping after initial loading, additional shims were inserted for the tie-plate with small load. Finally, pre-compress load of all tie-plate became within the target range. These results indicate that the pre-compression has been successfully completed using the developed jigs. In this paper, the design and analysis of the pre-compression jigs and the result of the pre-compression are reported.