International Thermonuclear Experimental Reactor Design Research Articles

Modeling of toroidal momentum transport induced by neoclassical toroidal viscosity torque for ITER scenarios

Neoclassical toroidal viscosity (NTV) torque caused by resonant magnetic perturbation (RMP) and the induced toroidal momentum transport are investigated for International Thermonuclear Experimental Reactor (ITER) scenarios through numerical modeling. The NTV torque is calculated using the NTVTOK code including the bounce-drift resonant effect, and the toroidal rotation evolution is modeled by solving a toroidal momentum transport equation that couples momentum source (NTV torque) and the momentum diffusion effect. The variation of RMP coil phasing (defined as toroidal phase difference between different rows of RMP coils) results in different types of plasma response and hence different features of NTV torque and toroidal momentum transport. The bounce-drift resonant effect enhances NTV torque and induces more significant toroidal rotation variation than simulations that adopt the bounce-averaged NTV model. With the initial rotation of the ITER design, plasma rotation is braked by NTV torque, but it may be sustained at moderate amplitude due to electron contributions to NTV torque. It is also found that initially static or slowly rotating plasma can be accelerated by NTV torque either toward co- Ip or counter- Ip ( Ip indicates plasma current) direction, indicating that NTV torque can be regarded as a momentum source for plasma with low torque injection; for instance, radio-frequency heated plasma.

Nuclear Response Analysis of TFC for CFETR-I Using an Optimized GVR Approach

The nuclear analysis of the toroidal field coil (TFC) has been performed using China Fusion Engineering Test Reactor (CFETR)-I model of 33.75-degree torus sector with a pitched Neutral Beam Interface port. A novel optimized global variance reduction (GVR) approach has been applied to Monte Carlo N-Particle Transport Code (MCNP) simulation process to improve parallel computing efficiency. The nuclear response in TFC includes the nuclear heat density, the integral nuclear heat, the absorbed dose, the fast neutron fluence and the total neutron fluence. Those data were calculated and compared with the International Thermonuclear Experimental Reactor (ITER) design limits. The results show that the fast neutron fluence, total neutron fluence and absorbed dose in some parts of inboard insulator exceed the ITER design limits, which means that more advanced shield materials should be used in radiation shielding design.

Controlled Fusion: Magnetic and Inertial, Promises and Pitfalls

As with biomass, hydro, solar and wind power, fusion power can also generate clean energy, using deuterium, an isotope of hydrogen, abundantly available in our oceans. Our sun uses hydrogen in a fusion process to generate power. It has been demonstrated that fusion power can be generated on earth, under carefully controlled conditions using deuterium and tritium instead of hydrogen. There are two fundamental approaches to controlled fusion: magnetic confinement fusion (MCF) first proposed at Princeton University in 1951, and inertial confinement fusion (ICF) that followed shortly thereafter, first proposed at the Lawrence Livermore Laboratories in 1970. Progress made on magnetic fusion led to the planning and construction of ITER (International Thermonuclear Experimental Reactor), expected to be completed in 2035. In this article, we explain the processes necessary to generate fusion power through MCF and ICF. Unlike nuclear power, as a practical means to generate electricity, controlled fusion has presented the technical/scientific community with a plethora of very difficult challenges. It is only recently, after decades of intense research in many laboratories worldwide, that we have begun to see devices being built on a fusion reactor scale and hence the design of ITER. The challenges are many but require patience and perseverance.

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.

A Study of the Consecutive Absorption/Desorption Cycles of ZrCo–H2System

Consecutive absorption/desorption cycles of the ZrCo–H2 system were studied to simulate the real International Thermonuclear Experimental Reactor (ITER) hydrogen getter system. A ZrCo getter was used in this paper instead of the depleted uranium (DU) getter material, which has been recently considered as the hydrogen getter in ITER. In a cyclic pressure-composition isotherm (PCI) measurement, the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo–H2 system in detail, especially for the desorption stage. This high-pressure PCI apparatus, however, shows cause and effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo–H2 system or in case of the DU–H2 system having multiple getter bed battery, a similar affection by previous getter status might be related and a similar aspect could be shown to consider further ITER design; for example, a need for control logic from PCI measurements using a high-pressure Sievert apparatus.

Manufacture and Acceptance Test of the Full Size ITER PF5 Conductor Sample

In the R&D process of the International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) coils, 16 m PF5 Cable-in-Conduit Conductor (CICC) was integrated at Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). The manufacture process of the qualification conductor was according to the ITER design. As required by the ITER Procurement Arrangement (PA), full size PF5 qualification conductor sample was manufactured and delivered to SULTAN facility (CRPP Villigen, Switzerland) for acceptance test. The PF5 conductor sample was made of two legs by one conductor bending, bottom hairpin box, upper terminations, support clamps, insulation, cryogenic piping and instrumentations. The concept of hairpin box and comb-type terminations configuration for ITER PF conductor sample which was co-contributed by ENEA and CEA was successfully verified by the first China PF2 conductor sample tested in SULTAN facility May 2010. The PF5 conductor sample test was initiated at the end of Feb. 2011. The test program included DC performance, cyclic loading, AC loss and MQE test. The conductor manufacture process, conductor sample assembly and test results were presented in this paper.

Methods and Instruments for Assessing the in-Core Gamma Heating of Materials

The reliable estimate gamma heating of materials in a research reactor core, including irradiated samples or core compartments, is recognized as directly related to reactor safety. A review of related reports indicates that the capability of accurate assessment of gamma heating of materials in nuclear reactor cores remains a matter of investigation. Numerous works, theoretical and/or experimental, have been performed in the frame of the design of the future fusion reactor ITER (International Thermonuclear Experimental Reactor) while relevant work is also reported concerning the future material testing reactor Jules Horowitz (RJH). In the present work, an extended review is presented about methods and instruments that have been developed in order to assess the heating of materials due to absorption of gamma radiation in a research reactor core. This compilation may contribute to define coherent measurement and computational strategies in order to reliably assess gamma heating.

Design of the Plasma Position and Shape Control in the ITER Tokamak Using In-Vessel Coils

The International Thermonuclear Experimental Reactor (ITER) is the next step toward the realization of electricity-producing fusion power plants, and it is planned to be in operation in 2016. ITER has been designed so as to reach the plasma burning condition and to operate with high-elongated unstable plasmas. However, due to the constraints that affect the machine realization, these open-loop unstable high-performance plasmas can hardly be stabilized using the poloidal field coils placed outside the tokamak vessel. For this reason, during the ITER design review phase, it has been proposed to investigate the possibility of using in-vessel coils in order to improve the best achievable performance of the vertical stabilization (VS) system. This paper proposes a new approach for the plasma current, position, and shape control design in the presence of in-vessel coils. Two control loops are designed: a first loop that guarantees the VS driving the voltage applied to in-vessel coils, and a second loop that controls the plasma current and up to 32 geometrical shape descriptors as close as possible to the reference values. The performance of the proposed control system is shown by means of simulations of some cases of interest.

Neutronics analysis of the international thermonuclear experimental reactor (ITER) MCNP “Benchmark CAD Model” with the ATTILA discrete ordinance code

There is currently extensive effort to develop and benchmark new codes to supplement the accepted computational tools for the nuclear design of ITER. The ITER management and quality program (MQP) officially requested benchmarking the newly developed FEM 3D, discrete ordinates code, ATTILA, to the results obtained by the CAD-based MCNP Monte Carlo codes. CAD-based computational tools are needed to reduce the turnaround time between changes in design and the subsequent nuclear analysis to key design parameters in an iterative process. There is currently extensive effort to develop and use a CAD–MCNP interface for design purposes to facilitate the modeling and analyses. ATTILA uses CAD-based geometrical input and as such it is a potentially quicker alternative to the MCNP code for the neutronics studies contingent on its success in predicting key neutronics parameters in large systems such as ITER. The ITER neutronics community had agreed to use a standard CAD model of ITER (40° sector, denoted as“Benchmark CAD Model”) to compare results for several responses selected for calculation benchmarking purposes to test the efficiency and accuracy of the CAD–MCNP approach developed by each ITER party in addition to ATTILA. In this paper we report such comparison to the results of the CAD-based MCNP developed at the University of Wisconsin for five selected nuclear responses. It is shown that ATTILA's results are within 25–30% of the MCNP results.

Mirror test for International Thermonuclear Experimental Reactor at the JET tokamak: An overview of the program

Metallic mirrors will be essential components of all optical spectroscopy and imaging systems for plasma diagnosis that will be used at the next-step magnetic fusion experiment, International Thermonuclear Experimental Reactor (ITER). Any change of the mirror performance, in particular, reflectivity, will influence the quality and reliability of detected signals. At the instigation of the ITER Design Team, a dedicated technical and experimental activity aiming at the assessment of mirror surface degradation as a result of exposure to the plasma has been initiated on the JET tokamak. This article provides a comprehensive overview of the mirror test program, including design details of the mirror samples and their supports, their locations within JET, and the issue of optical characterization of the mirrors both before and after exposure. The postexposure characterization is particularly challenging in JET as a consequence of an environment in which both tritium and beryllium are present.

国際熱核融合実験炉‘ITER’の工学設計活動から建設へ

The purpose of this paper is to explain, from the aspect of the superconducting coil system, the concept of ITER (International Thermonuclear Experimental Reactor), which is to be constructed in France under the international collaboration of the seven parties. The specifications and results of research and development on the superconducting model coils, which was performed during the Engineering Design Activities, were compared with the final design of ITER. The identification of the relevance of model coils compared to full-scale ITER coils and related limitations are discussed from the technical and project management point of view.

Simulation of cryogenic He spills as basis for planning of experimental campaign in the EVITA facility

Code validation activities have been promoted inside the European fusion development agreement (EFDA) to test the capability of codes in simulating accident phenomena in fusion facilities and, specifically, in the International thermonuclear experimental reactor (ITER). This work includes a comparison between three different computer codes (CONSEN, MAGS and MELCOR) and one analytical model (ITER Model) in simulating cryogenic helium releases into the vacuum vessel (VV) which contains hot structures. The scope was the evaluation of the transient pressure inside the VV. The results will be used to design a vent duct (equivalent diameter, length and roughness) to allow pressure relief for the protection of the VV, which has a maximum design pressure of 200 kPa. The model geometry is a simplified scheme preserving the main features of the ITER design. Based on the results of the simulations, a matrix of experiments was developed to validate the calculated results and to design the vent duct for the ITER VV. The experiments are planned to be performed in the EVITA test facility, located in the CEA Cadarache research centre (France).

Hybrid Scenario Development in DIII-D

Experiments in the DIII-D tokamak have demonstrated the ability to sustain ELMing H-mode discharges with high beta and good confinement quality under stationary conditions. These experiments have shown the ability to sustain normalized fusion performance (in terms of βNH89P /q952) at or above that projected for Qfus = 10 operation in the International Thermonuclear Experimental Reactor (ITER) design over a wide range in operating parameters. In the best cases, operation is maintained at the free boundary, n = 1 stability limit. Confinement is found to be better than standard H-mode confinement scalings over a wide range in operation space, and experimentally measured transport is consistent with predictions from the GLF23 transport code. Projections using the standard ITER H-mode scaling laws based on these discharges indicate that Qfus = 5 can be maintained for >5400 s in ITER at q95 = 4.5 while Qfus = 40 can be obtained for ~2400 s at q95 = 3.2. These projected performance levels further validate the ITER design and suggest that long-pulse, high neutron fluence operation as well as very high fusion gain operation may be possible in next-generation tokamaks.

Integrated Plasma Control in DIII-D

The integrated plasma control approach provides a systematic method for designing plasma control algorithms with high reliability and for confirming their performance off-line prior to experimental implementation. This approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance. Using this approach, required levels of robustness to model uncertainties and off-normal events can be quantified and incorporated in the design process. The DIII-D digital plasma control system (PCS) enables application of this method by providing a flexible programming environment and an architecture for real-time parallel operation of a set of computers that executes the large set of control algorithms needed for exploration of the advanced tokamak regime. The present work describes the DIII-D PCS and the approach, benefits, and progress made in integrated plasma control as applied to the DIII-D tokamak, with implications for the International Thermonuclear Experimental Reactor design and other next-generation tokamaks.

A low cost joint for the ITER PF coils, design and test results

The Poloidal Field (PF) coils of the International Thermonuclear Experimental Reactor (ITER) are designed using of NbTi cable-in-conduit conductors carrying a current up to 45 kA. Each PF coil consists of several modules wound in double pancakes. Joints connect that double pancakes electrically in series at the coil periphery. In order to minimize the manufacturing cost maintaining a high reliability, a low cost joint design was developed at CRPP and tested in SULTAN test facility in the frame of ITER task for PF coils. The concept of the joint is a single fully welded stainless steel box and one pitch joint through sectioned copper saddle blocks. The joint sample includes two full-scale joints, one in the hairpin configuration at the bottom of the sample and one as an overlap joint at mid length of the sample. Lifting the sample in the SULTAN facility, either the hairpin or the overlap joint can be placed in the center of the facility and tested under ITER-relevant operating conditions. The low cost joint design for the ITER PF coils is described and the test results (DC and AC loss) are presented in this paper.

DC Performances of ITER NbTi Conductors: Models vs. Measurements

Within the R&D program on the International Thermonuclear Experimental Reactor (ITER) Poloidal Field (PF) coils, a full size conductor sample was tested in the SULTAN facility (CRPP Villigen, Switzerland). This sample is composed of two straight ITER-like cable-in-conduit conductors, using the same NbTi strand. The two conductors are identical except that one leg has a cable containing steel wraps around the main subcables as in the ITER design, while the other has no wraps inside. The paper presents conductor DC test results compared to predictions given by various models developed within ITER-associated laboratories. These models aim to predict the DC behavior of the cable from the experimental performances of the single strand. They have to explain the observed voltage-current (V-I) or voltage-temperature (V-T) characteristics, including the thermal runaways. The lower experimental performances compared to all expectations have shown the necessity to revise the models and to introduce a possible uneven current distribution among the strands of the cables.

Analysis and optimization of the impact of ferromagnetic inserts on the toroidal field ripple in the International Thermonuclear Experimental Reactor

Perturbations of the axial symmetry of a toroidal magnetic field, or the toroidal field (TF) ripple, in tokamaks can potentially lead to losses of high-energy particles from the plasma. These losses can be significant and are one of the major concerns with respect to both the energy blance of the deuterium–tritium plasma and the heat load limits in the International Thermonuclear Experimental Reactor (ITER). In the ITER design, ferromagnetic inserts are used to reduce the TF ripple. In this work the impact of the ferromagnetic inserts on the TF ripple in the ITER is analysed to optimize their filling factor within the design constraints. The analysis of the high-mode spectrum of the amplitudes of ripple high modes is also carried out.

Development of ITER Divertor Vertical Target with Annular Flow Concept—I: Thermal-Hydraulic Characteristics of Annular Swirl Tube

Thermal-hydraulic tests for pressurized water in an annular tube with a twist fin have been performed to examine its applicability to high-heat-flux components of the International Thermonuclear Experimental Reactor (ITER) divertor. The annular swirl tube consists of two concentric tubes: an outer smooth tube and an inner tube with an external twist fin to enhance heat transfer of the cooling water in the annulus section between the outer and the inner tubes. Critical heat flux (CHF) tests under one-sided-heating conditions show that the annular swirl tube has as high removal limitation as the conventional swirl tube, the dimensions of which are similar to those of the outer tube of the annular swirl tube. A minimum axial velocity of 7.1 m/s is required for 28 MW/m2, the ITER design value. Pressure drops in the annulus section and the end return have been measured. The applicability of the existing correlations for heat transfer and CHF to the annular swirl tube has also been examined.

Design study for International Thermonuclear Experimental Reactor high resolution x-ray spectroscopy array

The impurity line and continuum emission for International Thermonuclear Experimental Reactor (ITER) reference H-mode and Internal Transport Barrier (ITB) plasmas were modeled using the SANCO impurity transport code. Using the instrument sensitivity for a spatially resolving crystal spectrometer array with doubly-curved crystals and two-dimensional detectors, signals, and signal-to-noise ratios were calculated for impurities including argon, iron, and krypton. These were shown to have lines suitable for the measurement of the ion temperature (0.5–30 keV) and the rotation over almost the entire plasma minor radius. The main contribution to the signal-to-noise is the plasma continuum radiation on which the lines are superimposed. The tolerable impurity concentration is limited by the incremental radiated power ΔPrad, there being a broad operating range between about 100 kW and 10 MW. The spectrometer array has now been integrated into the ITER design. A quasitomographic technique to reconstruct the Ti and rotation profiles, is reported by Ingesson et al. [C. Ingesson, these proceedings].

Fusion Research by Magnetic Confinement Devices in Japan (Plenary Paper, ICOPS 2003)

Remarkable progress in the past decade in tokamak magnetic confinement devices has made it possible to plan and design an experimental reactor, called the international thermonuclear experimental reactor (ITER), by the international collaboration. Data on plasma behavior obtained by the world-wide tokamak devices have been accumulated and analyzed in a coordinated manner which constitute a sound basis for the prediction of ITER core plasma performance. Large scale technology R&D including development, manufacturing, and testing of scalable models of major tokamak components has been also pursued that ensured technical feasibility of ITER design. Thus, ITER is now technically ready to start construction, and the governmental negotiation on its construction, operation, and exploitation activities, etc., has been underway since November 2001. After reaching agreement on important issues for ITER, the international ITER organization is expected to start in 2004, and the main purpose of its exploitation; control of significantly burning plasma will be achieved within a few decades. Aiming at the ultimate goal of steady-state operation at high plasma pressure for ITER and beyond, researches for advanced modes of plasma in tokamaks are continued by the existing experimental devices. Alternative approaches by nontokamak experimental devices in universities and national institutes have also been made in parallel. To reorganize these activities centered at ITER as a core program, the reform for more a efficient system of the entire Japanese fusion program were discussed in the special working group under the government (MEXT). This paper gives an overview of current activities and future program on magnetic confinement fusion research in Japan.