International Thermonuclear Experimental Reactor Operation Research Articles

Nonlinear Cone Model for Investigation of Runaway Electron Synchrotron Radiation Spot Shape

The runaway electron event is the fundamental physical phenomenon and tokamak is the most advanced conception of the plasma magnetic confinement. The energy of disruption generated runaway electrons can reach as high as tens of mega-electron-volt and they can cause a catastrophic damage of plasma-facing-component surfaces in large tokamaks and International Thermonuclear Experimental Reactor (ITER). Due to its importance, this phenomenon is being actively studied both theoretically and experimentally in leading thermonuclear fusion centers. Thus, effective monitoring of the runaway electrons is an important task. The synchrotron radiation diagnostic allows direct observation of such runaway electrons and an analysis of their parameters and promotes the safety operation of present-day large tokamaks and future ITER. In 1990 such diagnostic had demonstrated its effectiveness on the TEXTOR (Tokamak Experiment for Technology Oriented Research, Germany) tokamak for investigation of runaway electrons beam size, position, number, and maximum energy. Now this diagnostic is installed practically on all the present-day’s tokamaks. The parameter v┴/|v||| strongly influences on the runaway electron synchrotron radiation behavior (v|| is the longitudinal velocity, v┴ is the transverse velocity with respect to the magnetic field B). The paper is devoted to the theoretical investigation of runaway electron synchrotron radiation spot shape when this parameter is not small that corresponds to present-day tokamak experiments. The features of the relativistic electron motion in a tokamak are taken into account. The influence of the detector position on runaway electron synchrotron radiation data is discussed. Analysis carried out in the frame of the nonlinear cone model. In this model, the ultrarelativistic electrons emit radiation in the direction of their velocity v→ and the velocity vector runs along the surface of a cone whose axis is parallel to the magnetic field B. The case of the small parameter v┴/|v||| (v┴/|v|||<<1, linear cone model) was considered in the paper: Plasma Phys. Rep. 22, 535 (1996) and these theoretical results are used for experimental data analysis.

Results of All ITER TF Full-Size Joint Sample Tests in Japan

Nine toroidal field (TF) coils have been developed in Japan for the international thermonuclear experimental reactor (ITER). The joint resistance of TF coil should satisfy the requirement of smaller than 3 nano-ohm at 2 T of external magnetic field and 68 kA of transport current. Full-size joint sample (FSJS) tests were performed for joint development and for TF coil manufacture, as part of the process control. 11 FSJS tests are conducted in total. FSJS tests were conducted with assistance from a test faculty in the National Institute for Fusion Science as reported in a previous paper. All FSJS tests successfully satisfied the requirement of resistance less than 3 nΩ at 2 T. Additionally, the TF coil joints are subjected to cyclic electromagnetic force and warm-up/cool-down during the ITER operation. The authors investigated the joint performance for the abovementioned influence. The results showed no degradation in the joint resistance. Thus, the TF joint developed in Japan was qualified successfully.

TRIPOLI-4 simulation of the FNG copper benchmark experiment and the tritium production in the lithium diamond detector for ITER-TBM

Abstract For the future ITER (International Thermonuclear Experimental Reactor) operation needs, several D–T fusion neutron experiments have been conducted at FNG (Frascati Neutron Generator) facility. These FNG experimental results are useful to validate neutron transport codes and nuclear data libraries. The TRIPOLI-4 Monte-Carlo radiation transport code has been widely used on radiation shielding, criticality safety and fission reactor physics fields to support nuclear energy research and industrial applications. With the growing interest in using the TRIPOLI-4 code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the 14 MeV fusion neutronics calculations. Copper is a key material used in different components of ITER tokamak machine and the FNG copper benchmark experiment funded by F4E (Fusion for Energy) was recently introduced into the OECD/NEA SINBAD database. In the first part of this work, TRIPOLI-4 calculations were performed to simulate the D–T fusion neutron transport in a pure copper block of FNG-Cu experiments. The reaction rates were evaluated for various activation foils (93Nb(n,2n), 27Al(n,α), 58Ni(n,p), 115In(n,n’), 197Au(n,2n), 186W(n,γ), and 197Au(n,γ)) versus penetration depth in the Cu block. To study the measurement tool of tritium production for the future ITER Test Blanket Module (TBM), the second FNG experiment analyzed in this work arranged a thin layer of 6LiF (LiDia detector) placed inside a polyethylene cylinder and irradiated with FNG 14 MeV neutrons. TRIPOLI-4 simulation models were run to evaluate the tritium production in the LiDia detector. The purpose of this study is to investigate the TRIPOLI-4 calculations against the FNG experimental ones. TRIPOLI-4 calculation methods and results are reported.

Mechanical Stability of Filaments for ITER Diagnostic Pressure Gauges Relating to Creep and Fatigue

Diagnostic pressure gauges (DPGs) shall provide measurements of the neutral gas pressure in various locations of the International Thermonuclear Experimental Reactor (ITER) vacuum vessel. This parameter is essential for the basic control of ITER operation as well as for input to physics models of the plasma boundary. The hot cathode (filament) is the component of the DPG sensor that is exposed to most demanding loads. In order to reach the required electron emission from the filament, it has to be heated to high temperatures by a direct current. As a result, thermal stresses appear, and in the presence of a magnetic field, additional mechanical stresses caused by Lorentz forces arises. On the one hand, due to a large number of pulses foreseen in ITER, this load will be cyclic, which may result in filament failure caused by fatigue. On the other hand, the longest pulses in ITER are expected to be in the order of 30 min, and thus, filaments could fail due to creep. In order to verify that the filament of the DPG sensor can withstand fatigue and creep dedicated experiments have been conducted. Results of these tests are discussed in this article.

Toxicological Assessment of ITER-Like Tungsten Nanoparticles Using an In Vitro 3D Human Airway Epithelium Model.

The International Thermonuclear Experimental Reactor (ITER) is an international project aimed at the production of carbon-free energy through the use of thermonuclear fusion. During ITER operation, in case of a loss-of-vacuum-accident, tungsten nanoparticles (W-NPs) could potentially be released into the environment and induce occupational exposure via inhalation. W-NPs toxicity was evaluated on MucilAir™, a 3D in vitro cell model of the human airway epithelium. MucilAir™ was exposed for 24 h to metallic ITER-like milled W-NPs, tungstate (WO42−) and tungsten carbide cobalt particles alloy (WC-Co). Cytotoxicity and its reversibility were assessed using a kinetic mode up to 28 days after exposure. Epithelial tightness, metabolic activity and interleukin-8 release were also evaluated. Electron microscopy was performed to determine any morphological modification, while mass spectrometry allowed the quantification of W-NPs internalization and of W transfer through the MucilAir™. Our results underlined a decrease in barrier integrity, no effect on metabolic activity or cell viability and a transient increase in IL-8 secretion after exposure to ITER-like milled W-NPs. These effects were associated with W-transfer through the epithelium, but not with intracellular accumulation. We have shown that, under our experimental conditions, ITER-like milled W-NPs have a minor impact on the MucilAir™ in vitro model.

Current investigations on tritiated dust and its impact on tokamak safety

During International Thermonuclear Experimental Reactor operation, due to plasma–wall interaction, particles/dust will be created in sizes ranging from nanometers to tens of microns. The dust properties, especially their ability to be covered by a thin oxide electrostatic insulating layer, and surface topology deeply affect their tritium inventory. Consequently, physico-chemical properties specific to tritiated tungsten particles and consequence on particle behavior in the facility and environment must be carefully assessed. For size-relevant tungsten particles, the measured tritium inventory is ~10 GBq g−1. However, it varies with the particle specific surface area. Due to tritium beta decay and the oxide-insulating layer, dust exhibits a positive electrostatic self-charging. For a 5 µm particle in diameter with a 10 GBq g−1 tritium inventory, self-charging rate could lead to 5.5 104 elementary electric charges per day. These electrostatic properties could change the adhesion of dust on walls. In the case of a single particle, the adhesion will be reinforced due to image and dielectric forces. However, if the tritiated particle is part of an aggregate, the adhesion remains unknown. Due to the limited free path of the β emission in material, the tritium inventory carried by airborne particles cannot be measured in real time by conventional continuous radioactive aerosols monitors, and a new measurement strategy is needed for atmospheric surveillance in the workplace and of facility exhaust. Toxicity studies dealing with exposure to untritiated/tritiated tungsten particles of 100 nm have been undertaken. It was shown that these particles are rapidly dissolved in biologic media. Finally, after collection, dust must be confined to avoid its spreading into the environment. Different technical solutions are presented in this paper.

Plasma–tungsten interactions in experimental advanced superconducting tokamak (EAST)

Tungsten (W) is used as the armor material of the International Thermonuclear Experimental Reactor (ITER) divertor and is regarded as the potential first wall material of future fusion reactors. One of the key challenges for the successful application of W in fusion devices is effective control of W at an extremely low concentration in plasma. Understanding and control of W erosion are not only a prerequisite for W impurity control, but also vital concerns to plasma-facing component (PFC) lifetime. Since the application of ITER-like water-cooled full W divertor in EAST in 2014, great efforts were made to investigate W erosion by experiment and simulation. A spectroscopic system was developed to provide a real-time measurement of W sputtering source. Both experiment and simulation results indicate that carbon (C) is the dominant impurity causing W sputtering in L-mode plasmas, which comes from the erosion of C plasma-facing material (PFM) in the lower divertor and the main chamber limiters. The mixture layer on the surface of W PFCs formed through redeposition or the wall coating can effectively suppress W erosion. Increasing the plasma density and radiation can reduce incident ion energy, thus alleviating W sputtering. In H-mode plasmas, control of edge localized mode (ELM) via resonant magnetic perturbation (RMP) proves to be capable of suppressing intra-ELM W erosion. The experiences and lessons from the EAST W divertor are beneficial to the design, manufacturing and operation of ITER and beyond.

14 MeV neutron streaming calculations for JET-like maze entrance using TRIPOLI-4 Monte Carlo code

For the future ITER (International Thermonuclear Experimental Reactor) operation needs, D-T fusion experiments named DTE2 will be conducted in 2020 at JET (Joint European Torus). During the DTE2 operations, 14 MeV fusion neutrons will be generated. One of the aims of the DTE2 neutronics experiments is to investigate the D-T neutron streaming along the penetrations of biological shield of the JET Torus Hall. The TRIPOLI-4 Monte-Carlo radiation transport code has been extensively used on radiation shielding analyses. The purpose of this study is to evaluate the variance reduction (VR) techniques of TRIPOLI-4 on the 14 MeV neutron streaming calculations for a JET-like maze entrance. A JET-like torus hall building of 40 m x 40 m x 4.2 m and a maze entrance of 11.8 m length in 5-section configuration of concrete were first modeled. The standard INIPOND VR method and the new developed VR method AMS (Adaptive Multilevel Splitting) were then presented and discussed. Using the single-step approach combined with AMS techniques, the TRIPOLI-4 calculations results including neutron flux and ambient dose equivalent maps, the VR performance, and the user-friendly AMS input of TRIPOLI-4 code are reported.

Tritium supply and use: a key issue for the development of nuclear fusion energy

Full power operation of the International Thermonuclear Experimental Reactor (ITER) has been delayed and will now begin in 2035. Delays to the ITER schedule may affect the availability of tritium for subsequent fusion devices, as the global CANDU-type fission reactor fleet begins to phase out over the coming decades. This study provides an up to date account of future tritium availability by incorporating recent uncertainties over the life extension of the global CANDU fleet, as well as considering the potential impact of tritium demand by other fusion efforts. Despite the delays, our projections suggest that CANDU tritium remains sufficient to support the full operation of ITER. However, whether there is tritium available for a DEMO reactor following ITER is largely uncertain, and is subject to numerous uncontrollable externalities. Further tritium demand may come from any number of private sector “compact fusion” start-ups which have emerged in recent years, all of which aim to accelerate the development of fusion energy. If the associated technical challenges can be overcome, compact fusion programmes have the opportunity to use tritium over the next two decades whilst it is readily available, and before full power DT operation on ITER starts in 2035. Assuming a similar level of performance is achievable, a compact fusion development programme, using smaller reactors operating at lower fusion power, would require smaller quantities of tritium than the ITER programme, leaving sufficient tritium available for multiple concepts to be developed concurrently. The development of concurrent fusion concepts increases the chances of success, as it spreads the risk of failure. Additionally, if full tritium breeding capability is not expected to be demonstrated in DEMO until after 2050, an opportunity exists for compact fusion programmes to incorporate tritium breeding technology in nearer-term devices. DD start-up, which avoids the need for external tritium for reactor start-up, is dependent upon full tritium breeding capability, and may be essential for large-scale commercial roll-out of fusion energy. As such, from the standpoint of availability and use of external tritium, a compact route to fusion energy may be more advantageous, as it avoids longer-term complications and uncertainties in the future supply of tritium.

DiracR-matrix calculations for the electron-impact excitation of neutral tungsten providing noninvasive diagnostics for magnetic confinement fusion

Neutral tungsten is the primary candidate as a wall material in the divertor region of the International Thermonuclear Experimental Reactor (ITER). The efficient operation of ITER depends heavily on precise atomic physics calculations for the determination of reliable erosion diagnostics, helping to characterise the influx of tungsten impurities into the core plasma. The following paper presents detailed calculations of the atomic structure of neutral tungsten using the multiconfigurational Dirac-Fock method, drawing comparisons with experimental measurements where available, and includes a critical assessment of existing atomic structure data. We investigate the electron-impact excitation of neutral tungsten using the Dirac R-matrix method and, by employing collisional-radiative models, we benchmark our results with recent Compact Toroidal Hybrid measurements. The resulting comparisons highlight alternative diagnostic lines to the widely used 400.88nm line.

Radiation shielding calculations for a 3D ITER benchmark model using TRIPOLI-4® Monte Carlo code

The International Thermonuclear Experimental Reactor (ITER) is currently under construction in France. From the radiation shielding and radiation protection points of view, an intense and large neutron source with energy around 14.1 MeV will be generated from the D-T plasma zone during the ITER operation and diverse gamma-ray sources from neutron activation of the reactor structure materials and coolant should be considered for the reactor operation and maintenance. To decrease the radiation impacts caused by these neutron and gamma sources, iterative designs and nuclear analyses of ITER components are currently performed with three dimensional Monte Carlo radiation transport calculations. Due to the important dimensions of ITER, the thick tokamak blanket modules, and the diagnostic and functional port openings, variance reduction techniques are essential in these Monte Carlo neutron transport calculations. To verify the reactor components design models and to check the radiation transport calculation results, advanced graphic features of the calculation tool are also necessary. With the growing interest in using the TRIPOLI-4® Monte Carlo radiation transport code for ITER applications, the aim of this paper is to study the feasibility to use variance reduction features of TRIPOLI-4 code on a 3D ITER benchmark model which is a 40° toroidal segment including 5796 vol cells. The calculation results reported in this paper include the axial and radial profiles of the inboard TF coil heating and the neutron flux attenuation through the equatorial port plugs and shield. The performance of the TRIPOLI-4 graphic tool under its parallel computing mode was also evaluated.

A parametric study of helium retention in beryllium and its effect on deuterium retention

Beryllium samples have been exposed in the PISCES-B linear plasma device to conditions relevant to the International Thermonuclear Experimental Reactor (ITER) in pure He, D, and D/He mixed plasmas. Except at intermediate sample exposure temperatures (573–673 K) He addition to a D plasma is found to have a beneficial effect as it reduces the D retention in Be (up to ∼55%), although the mechanism is unclear. Retention of He is typically around 1020–1021 He m−2, and is affected primarily by the Be surface temperature during exposition, by the ion fluence at <500 K exposure, but not by the ion impact energy at 573 K. Contamination of the Be surface with high-Z elements from the mask of the sample holder in pure He plasmas is also observed under certain conditions, and leads to unexpectedly large He retention values, as well as changes in the surface morphology. An estimation of the tritium retention in the Be first wall of ITER is provided, being sufficiently low to allow a safe operation of ITER.

Synchronized fusion development considering physics, materials and heat transfer

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q = 10. ITER will demonstrate D–T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m−2 and an average heat flux of 0.1 MW m−2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like Wf/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Environmental dose rate assessment of ITER using the Monte Carlo method

Exposure to radiation is one of the main sources of risk to staff employed in reactor facilities. The staff of a tokamak is exposed to a wide range of neutrons and photons around the tokamak hall. The International Thermonuclear Experimental Reactor (ITER) is a nuclear fusion engineering project and the most advanced experimental tokamak in the world. From the radiobiological point of view, ITER dose rates assessment is particularly important. The aim of this study is the assessment of the amount of radiation in ITER during its normal operation in a radial direction from the plasma chamber to the tokamak hall. To achieve this goal, the ITER system and its components were simulated by the Monte Carlo method using the MCNPX 2.6.0 code. Furthermore, the equivalent dose rates of some radiosensitive organs of the human body were calculated by using the medical internal radiation dose phantom. Our study is based on the deuterium-tritium plasma burning by 14.1 MeV neutron production and also photon radiation due to neutron activation. As our results show, the total equivalent dose rate on the outside of the bioshield wall of the tokamak hall is about 1 mSv per year, which is less than the annual occupational dose rate limit during the normal operation of ITER. Also, equivalent dose rates of radiosensitive organs have shown that the maximum dose rate belongs to the kidney. The data may help calculate how long the staff can stay in such an environment, before the equivalent dose rates reach the whole-body dose limits.

Experiment and analysis of hypervapotron mock-ups for preparing the 2nd qualification of the ITER blanket first wall

According to the increased heat flux condition up to 5 MW/m 2 in the International Thermonuclear Experimental Reactor (ITER), new design of the blanket first wall (FW) has been considered and the analysis was performed with ANSYS-CFX for checking its temperature with the ITER operation conditions. And a semi-prototype of the FW was proposed to be tested with the similar heat flux conditions under the second qualification for the FW procurement. In order to investigate the fabrication procedure and analysis capability of the code, two types of mock-up were fabricated according to the current semi-prototype design except for bending shape; one with hypervapotron and another without it. They were tested with KoHLT-2 (Korea Heat Load Test) facility and the results were compared with the ones by CFX code. The mass flow rate of inlet coolant was the same as the ITER condition and heat flux was loaded up to 0.48 MW/m 2 heat flux. The results show that the temperature of the mock-up can be predicted using the CFX code even with the complex geometry and the hypervapotron shows its function to increase the cooling.

Flux and dose rate evaluation of iter system using MCNP5

The International Thermonuclear Experimental Reactor (ITER) will perform Deuterium-Tritium (DT) plasma experiments and the neutrons production rate at 14.1 MeV will achieve the level of 1013 n.cm-2.s-1. In this work, the neutron flux and the dose rate during ITER operation has been calculated using the one-dimensional model of the Monte Carlo code MCNP5 and the FENDL/MC-2.1 nuclear data library. The neutron flux and dose rate associated during normal ITER operation were determined along the radial machine direction. Calculations for two different types of concrete compositions were performed to investigate the impact of the bioshield filling materials on the dose rate estimation. The results show that the dose rate level near to the outer wall of the tokamak hall is close to the allowable limit dose. Taking into account the use of large boron concentration in the biological shield concrete (2.9% weight fraction), it was obtained a dose rate reduction by one order of magnitude. The dose rate is dominated by the secondary gamma ray. The magnitude of the dose rate on the outside hall of bioshield during normal ITER operation can not be considered low in accordance with the result found in the simulation performed in this work, i.e., 1 µSv.h-1.

Operational Radioactivity Evaluation of ITER Diagnostic Neutron Activation System

At the International Thermonuclear Experimental Reactor (ITER), a neutron activation system is foreseen to measure neutron fluence on the first wall and to evaluate total fusion power from ITER plasma. This system utilizes the method of counting gamma radiation from the metal sample irradiated by the neutron flux at the irradiation end located near the plasma. Evaluation of the operational radioactivity of the sample, which has been induced during ITER operation, is necessary in order to determine sample material, mass, and the appropriate irradiation time. Neutronic calculations are performed to evaluate operational activity of the sample. Activation of the sample is calculated according to the irradiation time for various sample materials such as Si, Al, Ti, Fe, Nb, and Cu and by assuming that the samples are irradiated at the midplane inboard region where the neutron flux is one of the strongest.

Sensitivity Study for the Optimization of the Viewing Chord Arrangement of the ITER Poloidal Polarimeter

In order to optimize the viewing chord arrangement of the poloidal polarimeter in International Thermonuclear Experimental Reactor (ITER), we studied the sensitivity between the toroidal current profile and the viewing chord position. The current profile is scanned around the ITER operation scenario which is estimated as one of burning plasma operations. The sensitive viewing chords to distinguish changes of the central toroidal current density, the finite beta effect and the magnetic shear, are shown. The physical mechanisms of those sensitivities are also shown.

Neutron Spectrometer for DD/DT Burning Ratio Measurement in Fusion Experimental Reactor

The most feasible fuels for a fusion reactor are D (Deuterium) and T (Tritium). DD and/or DT fusion reaction or nuclear burning reaction provides two kinds of neutrons, DD neutron and DT neutron, respectively. DD/DT burning ratio, which can be estimated by DD/DT neutron ratio in the burning plasma, is essential for burn control, alpha particle emission rate monitoring and tritium fuel cycle estimation. Here we propose a new neutron spectrometer for the absolute DD/DT burning ratio measurement. The system consists of a Proton Recoil Telescope (PRT) and a Time-of-Flight (TOF) technique. We have conducted preliminary experiments with a prototype detector and a DT neutron beam (ø20 mm) at the Fusion Neutronics Source, Japan Atomic Energy Agency (JAEA), to assess its basic performance. The detection efficiency obtained by the experiment is consistent with the calculation results in PRT, and sufficient energy resolution for the DD/DT neutron discrimination has been achieved in PRT and TOF. The validity of the Monte Carlo calculation has also been confirmed by comparing the experimental results with the calculation results. The design consideration of this system for use in ITER (International Thermonuclear Experimental Reactor) has shown that this system is capable of monitoring the line-integrated DD/DT burning ratio for the plasma core line of sight with the required measurement accuracy of 20% in the upper 4 decades of the ITER operation (fusion power: 100 kW-700 MW).

Neutron Spectrometer for DD/DT Burning Ratio Measurement in Fusion Experimental Reactor

The most feasible fuels for a fusion reactor are D (Deuterium) and T (Tritium). DD and/or DT fusion reaction or nuclear burning reaction provides two kinds of neutrons, DD neutron and DT neutron, respectively. DD/DT burning ratio, which can be estimated by DD/DT neutron ratio in the burning plasma, is essential for burn control, alpha particle emission rate monitoring and tritium fuel cycle estimation. Here we propose a new neutron spectrometer for the absolute DD/DT burning ratio measurement. The system consists of a Proton Recoil Telescope (PRT) and a Time-of-Flight (TOF) technique. We have conducted preliminary experiments with a prototype detector and a DT neutron beam (ø20 mm) at the Fusion Neutronics Source, Japan Atomic Energy Agency (JAEA), to assess its basic performance. The detection efficiency obtained by the experiment is consistent with the calculation results in PRT, and sufficient energy resolution for the DD/DT neutron discrimination has been achieved in PRT and TOF. The validity of the Monte Carlo calculation has also been confirmed by comparing the experimental results with the calculation results. The design consideration of this system for use in ITER (International Thermonuclear Experimental Reactor) has shown that this system is capable of monitoring the line-integrated DD/DT burning ratio for the plasma core line of sight with the required measurement accuracy of 20% in the upper 4 decades of the ITER operation (fusion power: 100 kW-700 MW).