Plasma Operation Scenario Research Articles

Development of pulsed plasma operation scenario and required conditions in JA DEMO

We have developed the pulsed plasma operation scenarios for JA DEMO, a design concept of the steady-state tokamak demonstration reactor, to clarify controls of the current profile and power required for the operation. We compare the scenarios when injecting electron cyclotron waves only and both neutral beam and electron cyclotron waves for external heating and current drive. We demonstrate current profile control that maintains the minimum value of the safety factor above one and avoids creating the local minima in the safety factor profile and power control by argon seeding that maintains the fusion power constant at the desired value and reduces the heat load on the divertor, performing long-time integrated modeling simulations. We clarify the conditions of the heating and current drive system and impurity injection system required for such control. The dependence of power control on argon anomalous transport coefficients is investigated. We have the prospect of maintaining the fusion power of 1 GW for more than two hours, i.e. obtaining the required plasma performance determined using a systems code.

Evaluation of the impact of plasma operation scenarios on the fusion reactor blanket design using an integrated numerical plasma and neutronics analysis suite

A very limited operation scenario has been used for fusion neutronics for the design of tritium breeding blankets. However, many advanced operation scenarios are under development and the impact of choosing a plasma operation scenario for fusion neutronics, which requires integration of neutronics analysis with plasma analysis, needs to be evaluated. An automated process from plasma analysis to neutronics analysis to create a viable fusion plasma neutron source was considered using a few different codes. In this work, we develop an integrated suite combining plasma modeling codes and neutronics codes. McCARD, a Monte-Carlo neutron-photon transport simulation code, is integrated into TRIASSIC, a tokamak reactor integrated automation suite for simulation and calculation, as the module for neutronics analysis of various plasma operation scenarios. McCARD integration allows TRIASSIC to perform neutronics analysis along with analytical and predictive plasma simulations. To evaluate this integrated suite, KSTAR experimental data were used to calculate the neutron flux at the first wall via TRIASSIC with the McCARD module in three different cases. First, the effect of plasma shape parameters, elongation and triangularity, on neutron flux is analyzed. Second, plasma scenario with the changes to the shape and the performance of plasma over time is studied to evaluate the time varying simulation. Third, two different plasma scenarios are compared to analyze the impact of plasma performance on neutron flux.

KSTAR ECH System Development Progress

The KSTAR ECH system has been upgraded to ensure steady-state advanced operation of KSTAR. The planned ECHs include a total of six 1 MW, 300 s ECH systems, of which four 105/140 GHz dual frequency systems have been commissioned and are being used in KSTAR experiments. Additionally, a 170 GHz Russian gyrotron has been commissioned and is ready for injection. The final unit, a multifrequency gyrotron similar to Japan’s ITER gyrotron, is scheduled for installation and operation in early 2025. To achieve the KSTAR mission of high-performance long-pulse operation using a plasma operation scenario with a high poloidal beta, ECH requires stable and long pulses, which typically results in the reduction of ECH power during campaigns. Consequently, one ECH unit with power ranging from 0.5 to 0.7 MW was utilized in the KSTAR experiments. The ECH power and launching angle can be controlled by the Plasma Control System (PCS) which maximizes the ECH effect at startup and at the flat top of the discharge. This paper describes the development status of the KSTAR ECH system and reports the longpulse operation results of the ECH system for a dummy load and KSTAR.

Parameter study of L–H transition for plasma operation scenario development in JA DEMO

We have investigated the L–H transition in JA DEMO, a design concept of the steady-state tokamak fusion demonstration reactor. We develop an L–H transition model, which determines the minimum pedestal growth time such that a power ratio, the net plasma loss power across the separatrix over the L–H threshold power, is maintained at a constant value during pedestal growth for a given external heating power. We conduct parameter studies using an integrated modeling code, considering impurity transport. The dependence of the heating power and time required for the L–H transition on the plasma density, temperature, threshold power, impurity density fraction, and impurity transport coefficients is clarified. We evaluate the appropriate impurity density fraction for the L–H transition and the divertor heat load reduction. The fraction depends on the impurity transport coefficients and external heating power. The obtained results are necessary to confirm the L–H transition feasibility and to develop the plasma operation scenario consistent with the various engineering constraints in a design-stage reactor. The heating power required for the L–H transition turns out to be lower than the power supposed to be installable in the current design activities of JA DEMO.

Neutral Beams for Neutron Generation in Fusion Neutron Sources

Neutral beam injection is supposed to be the main source of high-energy particles, driving non-inductive current and generating primary neutrons in fusion neutron sources design based on tokamaks. Numerical simulation of high-energy particles’ thermalization in plasma and fusion neutron emission is calculated by novel dedicated software (NESTOR code). The neutral beam is reproduced statistically by up to 109 injected particles. The beam efficiency and contribution to primary neutron generation is shown to be dependent on the injection energy, input current, and plasma temperature profile. A beam-driven plasma operation scenario, specific for FNS design, enables the fusion rate and neutron generation in plasma volume to be controlled by the beam parameters; the resultant primary neutron yield can be efficiently boosted in plasma maintained at a relatively low temperature when compared to ‘pure’ fusion reactors. NESTOR results are applicable to high-precision nuclear and power balance estimations, neutron power loads distribution among tokamak components, tritium generation in hybrid reactors, and for many other tasks critical for FNS design.

Density peaking in JET—determined by fuelling or transport?

Core density profile peaking and electron particle transport have been extensively studied by performing several dimensionless collisionality (υ*) scans with other matched dimensionless profiles in various plasma operation scenarios on the Joint European Torus (JET). This is the first time when electron particle transport coefficients in the H-mode have been measured on JET with high resolution diagnostics, and therefore we are in a position to distinguish between the neutral beam injection (NBI) source and inward electron particle pinch in contributing to core density peaking. The NBI particle source is found to contribute typically 50%–60% to the electron density peaking in JET H-mode plasmas where Te/Ti ~ 1 or smaller and at υ* = 0.1–0.5 (averaged between r/a = 0.3–0.8), and being independent of υ* within that range. In these H-mode plasmas, the electron particle transport coefficients, De and ve, are small, thus giving rise to the large influence of NBI fueling with respect to transport effect on peaking. In L-mode plasma conditions, the role of the NBI source is small, typically 10%–20%, and the electron particle transport coefficients are large. These dimensionless υ* scans give the best possible data for model validation. TGLF simulations are in good agreement with the experimental results with respect to the role of NBI particle source versus inward pinch in affecting density peaking, both for the H-mode and L-mode υ* scans. It predicts, similarly to experimental results, that typically about half of the peaking originates from the NBI fuelling in the H-mode and 10%–20% in the L-mode. GENE simulation results also support the key role of NBI fuelling in causing a peaked density profile in JET H-mode plasma (Te/Ti ~ 1 and υ* = 0.1–0.5) and, in fact, give an even higher weight on NBI fuelling than that experimentally observed or predicted by TGLF. For the non-fuelled H-mode plasma at higher Te/Ti = 1.5 and lower βN and υ*, both TGLF and GENE predict peaked density profiles, therefore agreeing well with experimental steady-state density peaking. Overall, the various modelling results give a fairly good confidence in using TGLF and GENE in predicting density peaking in quite a wide range of plasma conditions in JET.

Calculation method for pulsed magnetic field energy supplied to Nb3Sn ITER CS conductors during SULTAN stability tests

Cable-In-Conduit Conductors (CICCs) for the ITER Central Solenoid (CS) magnets are designed to operate in the presence of fast changing current and magnetic field during the plasma-operating scenario. For ITER, the AC loss of several types of Nb3Sn CICCs was experimentally tested, but only very limited experimental data is available for quantitative analysis of the minimum quench energy (MQE). In the SULTAN testing facility (Swiss Plasma Centre) few CS conductors were tested on MQE, but the magnetic field amplitude and ramp rate settings are far from the actual ITER operating conditions. Nevertheless, such tests are needed as a basis to calibrate and benchmark the codes that describe the quench behavior. Moreover, during the stability tests in Sultan, the temperature measurements show severe fluctuations, which can introduce a large error for the energy calculation. An interpretation is given for the temperature fluctuation and a procedure is proposed to significantly reduce the error in the pulsed energy calculation.

Progress in nuclear analyses of the ITER TBM Port Plug with Dummy TBMs

To achieve the validation and testing of tritium breeding blanket concepts, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three equatorial ports of the ITER tokamak. The TBM Port Plugs consist of Frame and TBM-Sets, which will be tested sequentially, each one tailored to the specific plasma operation scenario and test objective. In case a TBM-Set is not available, it can be replaced by a Dummy TBM at any time. TBM Frame and Dummy TBM are made of stainless steel and also need to meet ITER requirements on shielding, heat removal and safety, including activation. The maintenance operations for the TBM Port Plug include various activities in the Port Cell as well in the Hot Cell. In order to assess the requirements, a detailed MCNP model of the TBM Port Plug (PP), the Pipe Forest (PF) inside the Port Interspace (PI) and the Bioshield Plug (BP) was developed and integrated into the ITER neutronics “C-Model”.This analysis concerns the neutronic behaviour of the TBM PP with two Dummy TBMs by means of activation and shut-down dose rate (SDDR) calculations. The analysis includes the evaluation of the radiation fields around the TBM PP inside the ITER tokamak at several cooling times. Two configuration scenarios have been considered, one with a PF and PF structure inside the PI and one without this structure. The results show that the contribution to the SDDR from the TBM PP at 12 days after reactor shut down is small compared to the contribution originating from the PF structure.

Quench energy studies in ITER conductors for different magnetic field perturbations with Jackpot and THEA combined models

The electromagnetic–thermal models for the cable-in-conduit conductors (CICCs) JackPot-ACDC and THEA (thermal, hydraulic and electric analysis of superconducting cables) are combined to predict the stability of ITER central solenoid conductors. The combination of both models allows the prediction of the effect of any type of magnetic field perturbation in time, relevant for the magnet coils during the plasma operation scenario of the reactor. At present, there is no experiment for testing the stability of the ITER Nb3Sn conductors under such conditions. Only limited experimental data on minimum quench energy (MQE), defining the conductor stability, are available but the time and magnetic field amplitude settings are completely different from the actual ITER operating conditions. Nevertheless, such tests are useful as a basis to calibrate and benchmark the codes. The JackPot-THEA combination allows us to determine the MQE for any magnetic field change in time and to fully describe the involved electromagnetic phenomena in strand-level detail in terms of local power dissipation and (peak) electric field along all strands. Thermally, the computation is still on a global scale for identifying the quench initiation and propagation. The predictions from the combined codes are in good agreement with the experimental results and provide a solid basis for extrapolative scaling of the stability of CICCs under plasma operating conditions.

Core plasma design of the compact helical reactor with a consideration of the equipartition effect

Integrated physics analysis of plasma operation scenario of the compact helical reactor FFHR-c1 has been conducted. The DPE method, which predicts radial profiles in a reactor by direct extrapolation from the reference experimental data, has been extended to implement the equipartition effect. Close investigation of the plasma operation regime has been conducted and a candidate plasma operation point of FFHR-c1 has been identified within the parameter regime that has already been confirmed in LHD experiment in view of MHD equilibrium, MHD stability and neoclassical transport.

Status of CEA Magnet Design Tools and Applications to EU DEMO PF and CS Magnets

In the framework of the design activities conducted in the EU for dimensioning the future fusion demonstration reactor (DEMO), extensive analyses were carried out within the framework of the EUROfusion consortium, with the objective of defining the design of the DEMO magnets system. To this aim, CEA has developed ad-hoc predimensioning tools and associated methods in order to size the different magnets: toroidal field coils, central solenoid (CS) and poloidal field (PF) coils. Once predimensioned, the magnet concepts undergo a detailed evaluation procedure with more complex tools and methods. As these detailed analyses are time consuming, the whole design process benefits from an accurate and robust predimensioning process. The predimensioning tools described herein address various aspects driving the operational limits of the magnets while energized in the tokamak, related to electromagnetic, thermal, and mechanic phenomena. In this paper, we expose the latest developments in the predimensioning tools and the methods employed for obtaining a rapid and reliable macroscopic semi-analytical representation of the magnets. The assessment of the predimensioning tools on reference configurations (e.g., ITER) is also described. The application of these design tools on the DEMO configuration issued by EUROfusion is presented, and the resulting magnet design proposals are summarized: First, the PF system, composed of six coils, with one PF selected for study here. Second, the CS, with a modular geometry (five modules) and pancake winding. The approach here assesses the resilience of the CS to a fast-transient (breakdown). The outcomes of a sensitivity study on parameters/criteria are discussed, and extended to some tentative recommendations on the design approach for the DEMO PF and CS magnets. For the PF, a design is obtained for the PF5 coil following a dimensioning methodology consistent with the ITER PF system design with the same level of definition of the plasma operating scenario. As for the CS, it is stressed that consideration of the internal field (from shielding) substantially affects temperature margin results.

Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability margin of magnets. In the last decade ITER has launched a campaign for characterization of several types of NbTi and Nb3Sn CICCs comprising quench tests with a singular sine wave fast magnetic field pulse and relatively small amplitude. The stability tests, performed in the SULTAN facility, were reproduced and analyzed using two codes: JackPot-AC/DC, an electromagnetic-thermal numerical model for CICCs, developed at the University of Twente (van Lanen and Nijhuis 2010 Cryogenics 50 139–148) and multi-constant-model (MCM) (Turck and Zani 2010 Cryogenics 50 443–9), an analytical model for CICCs coupling losses. The outputs of both codes were combined with thermal, hydraulic and electric analysis of superconducting cables to predict the minimum quench energy (MQE) (Bottura et al 2000 Cryogenics 40 617–26). The experimental AC loss results were used to calibrate the JackPot and MCM models and to reproduce the energy deposited in the cable during an MQE test. The agreement between experiments and models confirm a good comprehension of the various CICCs thermal and electromagnetic phenomena. The differences between the analytical MCM and numerical JackPot approaches are discussed. The results provide a good basis for further investigation of CICC stability under plasma scenario conditions using magnetic field pulses with lower ramp rate and higher amplitude.

Development of a real-time simulation tool towards self-consistent scenario of plasma start-up and sustainment on helical fusion reactor FFHR-d1

This study closely investigates the plasma operation scenario for the LHD-type helical reactor FFHR-d1 in view of MHD equilibrium/stability, neoclassical transport, alpha energy loss and impurity effect. In 1D calculation code that reproduces the typical pellet discharges in LHD experiments, we identify a self-consistent solution of the plasma operation scenario which achieves steady-state sustainment of the burning plasma with a fusion gain of Q ~ 10 was found within the operation regime that has been already confirmed in LHD experiment. The developed calculation tool enables systematic analysis of the operation regime in real time.

Higher resolution helium measuring system for deuterium plasma on EAST tokamak via normal Penning gauge

Although the deuterium and helium have almost the same mass, a Penning Optical Gas Analyzer (POGA) system on the basis of the spectroscopic method and Penning discharging has been designed on EAST, since 2014. The POGA system was developed successfully in 2015, it was the first time that EAST could detect helium partial pressure in deuterium plasma (wall conditioning and plasma operation scenario). With dedicated calibration and proper adjustment of the parameters, the minimum concentration of helium in deuterium gas can be measured as about 0.5% instead of 1% on the other tokamak devices. Moreover, the He and D2 partial pressures are measured simultaneously. At present, the measurable range of deuterium partial pressure is 1 × 10−7 mbar to 1 × 10−5 mbar, meanwhile the range of helium is 1 × 10−8 mbar to 1 × 10−5 mbar. The measurable range can be modified by means of the adjustment of POGA system’s parameters. It is possible to detect the interesting part of the gas with a time resolution of less than 5 ms (the 200 ms because of conductance of transfer pipe at present). The POGA system was routinely employed to wall conditioning and helium enrichment investigation in 2015. Last but not the least, the low temperature plasma of POGA is generated by normal penning gauge Pfeiffer IKR gauge instead of Alcatel CF2P, which has been suspended for a few years and was used for almost all the POGA systems in the world.

Thermal behaviour analysis on ITER component cooling water system loop 2B

ITER cooling water system is composed by several cooling loops, the primary heat transfer loops that form the Tokamak Cooling Water System (TCWS), the secondary heat transfer loops that form the Component Cooling Water System (CCWS) and the Chilled Water System (CHWS) and a tertiary heat transfer loop which is the Heat Rejection System (HRS). The CCWS is further divided into CCWS-1, CCWS-2A, CCWS-2B, CCWS-2C, CCWS-2D depending on the water chemistry needs of clients and wetted area material. The component cooling water system loop 2B (CCWS-2B) has the function to remove heat load from coil power supply component, Neutral Beam Injectors (NBIs) system component and diagnostic system which are located in different buildings. As the total number of the client connections for the loop is a few hundreds, simplified thermal hydraulic analysis model has been developed to perform thermal analysis on the component cooling water system loop 2B. The curve of the cooling water temperature at client inlet and outlet during one cycle of the most demanding plasma operation scenario was obtained and the cooling water flow rate can meet the thermal removal requirement of client was also confirmed from this analysis. In addition, operation behaviour of the main heat exchanger for CCWS-2B in this thermal analysis was depicted for main heat exchanger selection purposes. This study has been carried out with the AFT Fathom code.

Design concept of K-DEMO for near-term implementation

A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. After the thorough 0D system analysis, the parameters of the main machine characterized by the major and minor radii of 6.8 and 2.1 m, respectively, were chosen for further study. The analyses of heating and current drives were performed for the development of the plasma operation scenarios. Preliminary results on lower hybrid and neutral beam current drive are included herein. A high performance Nb3Sn-based superconducting conductor is adopted, providing a peak magnetic field approaching 16 T with the magnetic field at the plasma centre above 7 T. Pressurized water is the prominent choice for the main coolant of K-DEMO when the balance of plant development details is considered. The blanket system adopts a ceramic pebble type breeder. Considering plasma performance, a double-null divertor is the reference configuration choice of K-DEMO. For a high availability operation, K-DEMO incorporates a design with vertical maintenance. A design concept for K-DEMO is presented together with the preliminary design parameters.

Fibrillation atriale

Active closed-loop control of the plasma safety factor profile (q-profile) and internal energy dynamics in nuclear fusion tokamak devices has the potential to significantly impact the success of the ITER project. These plasma properties are related to both the stability and performance of a given plasma operating scenario. In this work, we develop integrated feedback control algorithms to control the q-profile and internal energy dynamics in DIII-D advanced tokamak (high performance) scenarios. The feedback controllers are synthesized by embedding a nonlinear, physics-based, control-oriented partial differential equation model of the plasma dynamics into the control design and to be robust to uncertainties in the plasma electron density, electron temperature, and plasma resistivity profiles. The auxiliary heating and current-drive system and the total plasma current are the actuators utilized by the feedback controllers to control the plasma dynamics. Finally, the feedback controllers are tested both through simulations based on the physics-based model and experimentally in the DIII-D tokamak.

Experimental and Simulation Testing of Physics-model-based Safety Factor Profile and Internal Energy Feedback Controllers in DIII-D Advanced Tokamak Scenarios

Active closed-loop control of the plasma safety factor profile (q-profile) and internal energy dynamics in nuclear fusion tokamak devices has the potential to significantly impact the success of the ITER project. These plasma properties are related to both the stability and performance of a given plasma operating scenario. In this work, we develop integrated feedback control algorithms to control the q-profile and internal energy dynamics in DIII-D advanced tokamak (high performance) scenarios. The feedback controllers are synthesized by embedding a nonlinear, physics-based, control-oriented partial differential equation model of the plasma dynamics into the control design and to be robust to uncertainties in the plasma electron density, electron temperature, and plasma resistivity profiles. The auxiliary heating and current-drive system and the total plasma current are the actuators utilized by the feedback controllers to control the plasma dynamics. Finally, the feedback controllers are tested both through simulations based on the physics-based model and experimentally in the DIII-D tokamak.

Time parallelization of advanced operation scenario simulations of ITER plasma

This work demonstrates that simulations of advanced burning plasma operation scenarios can be successfully parallelized in time using the parareal algorithm. CORSICA -an advanced operation scenario code for tokamak plasmas is used as a test case. This is a unique application since the parareal algorithm has so far been applied to relatively much simpler systems except for the case of turbulence. In the present application, a computational gain of an order of magnitude has been achieved which is extremely promising. A successful implementation of the Parareal algorithm to codes like CORSICA ushers in the possibility of time efficient simulations of ITER plasmas.

Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″=0.3MW/m2) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.