Tokamak Reactor Studies Research Articles

Necessary and sufficient conditions for practical fusion power

Editors’ note: We invited Steven Cowley, former CEO of the UK Atomic Energy Authority, to comment on points raised by Robert Hirsch.Cowley replies: Undoubtedly, tokamaks have yielded by far the best plasma confinement of all fusion experiments. Indeed, the Tokamak Fusion Test Reactor at the Princeton Plasma Physics Laboratory in New Jersey and the Joint European Torus (JET) at the Culham Centre for Fusion Energy in the UK have achieved stable fusion conditions and significant fusion power—up to 16 MW in JET—from the deuterium–tritium reaction. Furthermore, detailed modeling from models validated against experimental data predicts that the international tokamak experiment ITER will attain a fusion “burn,” a state in which external heating is negligible and self-heating by the fusion-generated alpha particles is sufficient or almost sufficient to sustain the discharge.A burn would be the long-awaited scientific demonstration that energy production from fusion is possible. Only ITER offers the chance of reaching that hugely important milestone in the next two decades. However, as Robert Hirsch indicates, ITER will not prove the economic viability of fusion power. Such a determination is nontrivial, and without further R&D it is necessarily uncertain.Hirsch is wrong that tokamak reactor studies have ended in most parts of the world. For example, at the time of writing, demonstration tokamak reactor designs are being developed in the European Union (EU),11. G. Federici et al., Fusion Eng. Des. 109–111, 1464 (2016). https://doi.org/10.1016/j.fusengdes.2015.11.050 South Korea,22. K. Kim et al., Nucl. Fusion 55, 053027 (2015). https://doi.org/10.1088/0029-5515/55/5/053027 and China,33. B. N. Wan et al., IEEE Trans. Plasma Sci. 42, 495 (2014). https://doi.org/10.1109/TPS.2013.2296939 and less directed reactor studies are being pursued by all other ITER partners. Those studies address the well-known and serious technical issues raised by Hirsch. The authors made no attempt to downplay their significance. To appreciate the depth of the analysis, one has to read the extensive literature. I can only summarize briefly the current understanding of each of Hirsch’s issues.In fission and in fusion, cost is determined by much more than the mass of the core. Detailed estimates of the cost of electricity from the 2006 EU fusion reactor designs put the range44. D. Maisonnier et al., Fusion Eng. Des. 81, 1123 (2006). https://doi.org/10.1016/j.fusengdes.2005.08.055 between 0.03 and 0.09 €/kWh. ITER’s cost overruns, which are expected to be significantly less than Hirsch’s estimate, reflect a project that requires extensive R&D at every stage. They do not reflect the intrinsic industrial cost of components. Nonetheless, it is important to understand the ITER costs much better. Recent research, such as on the suppression of plasma turbulence, and expected improvements in technology, such as for superconducting magnets,55. B. N. Sorbom et al., Fusion Eng. Des. 100, 378 (2015). https://doi.org/10.1016/j.fusengdes.2015.07.008 suggest that innovation will drive down the cost and scale of tokamak reactors. Although I would not take any cost estimates too seriously, they indicate that tokamaks may enter the market in the right cost range. It is simply too early to be conclusive about cost.Hirsch is correct in identifying the quenching of superconducting magnets as being an issue for nuclear regulators. In fact, it is an issue with the French nuclear regulator for ITER. Technical studies of ITER show that a rapid quench of the superconducting magnets, caused by impact or otherwise, would not breach the containment of the vacuum vessel, let alone the main containment of the cryostat. Thus such an accident, although costly, would not endanger the surrounding population.The radioactivity of DT fusion reactors is a well-known issue.44. D. Maisonnier et al., Fusion Eng. Des. 81, 1123 (2006). https://doi.org/10.1016/j.fusengdes.2005.08.055 Material scientists have developed low-activation steels that reduce key impurities—nickel, for example—so that the radio isotopes produced by neutron bombardment are short-lived. With such materials, the activated material made in a fusion power plant will be low-level waste after 100 years.Tokamak reactors also face challenges not mentioned by Hirsch: tritium breeding and storage, for example.Success is not assured, but it is far too early to say that tokamaks fail against the Electric Power Research Institute criteria. Stimulating innovation on a broader range of ideas is also desirable. But we have an opportunity with ITER to create a burning plasma with an output of approximately 500 MW of fusion power. That opportunity should not be missed.ReferencesSection:ChooseTop of pageReferences <<CITING ARTICLES1. G. Federici et al., Fusion Eng. Des. 109–111, 1464 (2016). https://doi.org/10.1016/j.fusengdes.2015.11.050, Google ScholarCrossref, ISI2. K. Kim et al., Nucl. Fusion 55, 053027 (2015). https://doi.org/10.1088/0029-5515/55/5/053027, Google ScholarCrossref, ISI3. B. N. Wan et al., IEEE Trans. Plasma Sci. 42, 495 (2014). https://doi.org/10.1109/TPS.2013.2296939, Google ScholarCrossref, ISI4. D. Maisonnier et al., Fusion Eng. Des. 81, 1123 (2006). https://doi.org/10.1016/j.fusengdes.2005.08.055, Google ScholarCrossref, ISI5. B. N. Sorbom et al., Fusion Eng. Des. 100, 378 (2015). https://doi.org/10.1016/j.fusengdes.2015.07.008, Google ScholarCrossref, ISI© 2017 American Institute of Physics.

Relativistic ray-tracing of electron Bernstein waves in a spherical tokamak reactor

Electron Bernstein waves (EBW) can propagate between electron cyclotron harmonics in hot, dense plasmas with no high-density cutoff (unlike electron cyclotron waves). ARIES-ST is a spherical tokamak (ST) reactor study. A promising method of heating and driving current in such an ST reactor would be with electron Bernstein waves. These could be mode converted from microwave vacuum modes (X-mode and O-mode) launched from the edge. In this paper, fully relativistic ray-tracing calculations (both damping and propagation) of EBWs in ARIES-ST are presented, using the relativistic dispersion solver R2D2 and the ray-tracing code GENRAY. The ray paths, damping locations, and polarizations are compared with the more commonly used non-relativistic EBW ray-tracing. A range of frequencies are shown for EBWs with small n∥ that could be produced by the X-B mode conversion process, and EBWs with large n∥ ≈ 0.6, produced by the O-X-B mode conversion process. With the density and temperature profile chosen for this paper, the greatest depth into the core that could be reached with mode-converted EBWs is a radial location of approximately ρ = 0.4. Although the radial location of the damping in most cases was not significantly different between the relativistic and non-relativistic cases, there are differences in the poloidal locations, as well as in the polarizations of the wave along the ray path, especially in the damping region.

Overview of the Japanese fusion reactor studies programme

Fusion reactor design studies conducted in Japan since the fifth IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology in 1993 are introduced. During the 5 years since the last IAEA-TCM, the fusion reactor studies in Japan became more focused concentrating on tokamaks, helical devices and laser fusion. There are four tokamak reactor studies, namely, A-SSTR, CREST, DREAM and IDLT. There are helical reactor studies such as FFHR and one laser fusion reactor study on KOYO. Compared with the situation of 5 years ago having many types of confinement schemes being studied, the confinement schemes are now limited to three and each of the effort is larger involving greater number of scientists. The cooperation and exchanges between the design groups have increased through various committees, meetings and symposia.

Tritium release from lithium titanate, a low-activation tritium breeding material

The goals for fusion power are to produce energy in as safe, economical, and environmentally benign a manner as possible. To ensure environmentally sound operation low-activation materials should be used where feasible. The ARIES Tokamak Reactor Study has based reactor designs on the concept of using low-activation materials throughout the fusion reactor. For the tritium breeding blanket, the choices for low activation tritium breeding materials are limited. Lithium titanate is an alternative low-activation ceramic material for use in the tritium breeding blanket. To date, very little work has been done on characterizing the tritium release for lithium titanate. We have thus performed laboratory studies of tritium release from irradiated lithium titanate. The results indicate that tritium is easily removed from lithium titanate at temperatures as low as 600 K. The method of titanate preparation was found to affect the tritium release, and the addition of 0.1% H 2 to the helium purge gas did not improve tritium recovery.

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Divertor Engineering Studies for the International Tokamak Reactor

AbstractDuring the course of the International Tokamak Reactor studies, four interrelated issues for impurity control have been identified:1. pumping capability2. sputtering erosion3. heat removal4. lifetime.Recent plasma physics experiments and calculations indicate that the divertor is likely to operate with presheath plasma temperatures of only 10 to 20 eV. Surface sputtering erosion should be virtually eliminated under these conditions resulting in extended sputtering lifetimes for the divertor collector plates. The reduced sputtering means that the collector plates can be designed with only a thin layer of protection material facing the plasma, which should lead to reduced thermal stresses, greater heat load capability, and longer lifetimes. To assess the advantages and disadvantages of this type of divertor system, engineering trade-off studies have been performed. The studies include analyses of the materials, fabrication, thermomechanical behavior, disruptions, erosion/redeposition, tritium permea...

Parameter and cost optimizations for a modular stellarator reactor

The physical scaling and cost scaling of a modular stellarator reactor are described. It is shown that configurations based onl=2 are best able to support adequate beta, and physical relationships are derived which enable the geometry and parameters of anl=2 modular stellarator to be defined. A cost scaling for the components of the nuclear island is developed using Starfire (tokamak reactor study) engineering as a basis. It is shown that for minimum cost the stellarator should be of small aspect ratio. For a 4000 MWth plant, as Starfire, the optimum configuration is a 15 coil, 3 field period,l=2 device with a major radius of 16 m and a plasma minor radius of 2 m; and with a conservative wall loading of 2 MW/m2 and an average beta of 3.9%; the estimated cost per kilowatt (electrical) is marginally (7%) greater than Starfire.

Tokamak reactor studies

This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, nondivertor particle control, and improved reactor safety features.

Neutral-beam energy and power requirements for expanding-radius and full-bore start-up of tokamak reactors

Neutral-beam power arid energy requirements are compared for full-density, full-bore and expanding-radius start-up scenarios in an elongated plasma, The Next Step (TNS), as a function of beam pulse time and plasma density. Because of the similarity of parameters, the results should also be applicable to Engineering Test Facility (ETF) and International Tokamak Reactor (INTOR) studies. A transport model consisting of neoclassical ion conduction and anomalous electron conduction and diffusion based on ‘ALCATOR scaling’ leads to average densities in the range ⟨n⟩ ∼ 0.8–1.2 × 1014 cm−3 being sufficient for ignition. Neutral-deuterium-beam energies in the range 120–150 keV are adequate for penetration, with the required power injected into the plasma decreasing with increasing beam energy. The neutral-beam power decreases strongly with increasing beam pulse length tb until tb exceeds a few total energy confinement times, yielding tb ∼ 4–6 s for the TNS plasma. In addition to avoiding skin current effects and possibly allowing for a more impurity-free plasma initiation, the expanding-radius scenario has slightly reduced beam energy and/or power requirements. When the expanding-radius scenario is extended to even larger power reactors, a neutral-deuterium-beam energy of 150 keV remains sufficient for penetration.