Latest International Tokamak Project Plans Plasma Net Gain Goals in 2039

The latest (sept 2018) ITER Research Plan within the Staged Approach (Level III – Provisional Version) plans for plasma net gain goals to be achieved in 2039.

What is Q?

What then is the meaning of Q?

Quantitatively, Q is the out-versus-in power amplification ratio of the fusion reaction: the ratio of the amount of thermal power produced by hydrogen fusion compared to the amount of thermal power injected to superheat the plasma and initiate the reaction. ITER is designed to produce plasmas having Q over 10: meaning that injecting 50 megawatts of heating power into the plasma will produce a fusion output of at least 500 megawatts.

New Energy Times indicates that ITER will need 300 megawatts to operate. 500 megawatts might come out as heat.

Plasma energy breakeven, or Q=1, has never been achieved in a fusion device: the current record is held by the European tokamak JET (UK), which succeeded in generating a Q of 0.67. ITER’s Q value of over 10 makes it a first-of-kind machine
.
How did ITER’s designers choose the specific Q value? Accounting for the size of ITER’s vacuum vessel (830 cubic metres) and the strength of the confining magnetic field (5.3 Tesla), the ITER plasma can carry a current of up to 15 megaamperes. Under these conditions, an input thermal power of 50 megawatts is needed to bring the hydrogen plasma in the vessel to about 150 million degrees Celsius. This temperature in turn translates to a high enough velocity, among a sufficient population of hydrogen nuclei, to induce fusion at a rate that will produce at least 500 megawatts of thermal power output.

Why stop at a Q of 10? Why not design ITER for a Q of 30, or 50? The answer is clear: expense. For tokamaks, size and magnetic field strength matter. In simple terms, increasing Q would require an increase in the major radius or in the magnetic field strength. Either approach would have increased the cost of the device unnecessarily, whereas the achievement of Q ≥ 10 is sufficient to allow the primary scientific and technology goals of the project to be satisfied.

And a related question: Why not design ITER to produce electricity? This would also have required an increase in cost with no great benefit to the goals of the project. ITER is an experimental device designed to operate with a wide range of plasma conditions in order to develop a deeper understanding of the physics of burning plasmas, and to allow the exploration of optimum parameters for plasma operation in a power plant. The addition of the systems required to convert fusion power to high temperature steam to drive an electricity generator would not have been cost-effective, since the pattern of experimental operation of a tokamak such as ITER will allow for very limited generation of electricity.

Commercial fusion plants will be designed based on a power balance that accounts for the entire facility: the electricity output, sent to the industrial grid, compared to the electricity consumed by the facility itself—not only in tokamak heating, but also in secondary systems such as the electricity used to power the electromagnets, cool the cryogenics plant, and run diagnostics and control systems.

There are many possible risks at every step as they try to start experiments and then scale up.

2.6.4.3.1 Risk to the optimization of fusion power in DT plasma scenarios towards 15 MA/5.3 Tesla Q = 10 (~ 50 s) demonstration

By this stage of the operational plan many of the general risks to the achievement of the ITER Q = 10 goal may materialize. These are general to the project goals and are discussed in Appendix J. Below we discuss the specific risks that can affect the operational plan in this phase:
* H-mode confinement insufficient when integrated with ELM control by 3-D fields with invessel coils (plus pellet pacing) to achieve Q = 10 or ELM control capability is insufficient for 15 MA due to hardware limitations. This would require a reformulation of the research program above the plasma current level on which either of these two issues is identified.

Possible mitigation strategies would consist of:
a) switching to the hybrid/advanced scenario development at this stage and proceed to its optimization for Q over 5 and/or
b) to proceed to develop other high confinement regimes ELM-less regimes such as QH-mode or I-mode at this stage.

* Predictive and interpretative modeling of particle and thermal transport through the core to the edge, and of alpha-particle behavior required to develop the scenarios and burn control within them are not mature enough. This will require the progress of the experimental program to be based mostly on an empirical basis and, thus, slower progress of the research as the steps in Ip/Bt
, additional heating level, T-fraction, etc., may have to be smaller than strictly required in order to ensure that the plasma in the scenarios remain integrated with the operational/hardware constraints and MHD stable to ensure low disruptivity.

* Tritium throughput provided by the T-Plant not sufficient at the beginning of the experimental phase. Although it is expected that the T plant will be able to provide the required throughput to sustain Q = 10, ~50 s operation by the end of this operational phase, it may not be able to provide the required throughput at the beginning when 9.5 MA and 12.5 MA DT plasmas with high T-fraction are foreseen to be explored.

* Nuclear heating and/or AC losses of the superconducting coils are found to be much larger than expectations.

* Flux swing provided by the CS/PF system marginal for H-mode operation at 15 MA.

Research Program Accompanying Construction
5.1 H-mode issues

The research in present and near future tokamak experiments (ASDEX-Upgrade, DIII-D, EAST, JET, JT-60SA, KSTAR, WEST, etc.) should be focussed towards H-mode research in plasmas with the specific features of the ITER H-mode plasmas, in order to understand and develop validated models that can be used to predict plasma behaviour in H-mode scenarios in ITER. This involves not only the study of stationary conditions but also of the access/exit phases to/from H-modes, where plasma behaviour is already more difficult to control in present experiments, particularly with W PFCs. In ITER this is further complicated by the relatively low margin of Psep/PL-H and the need to maintain ELM control in these transient phases.

5.4.5 Impact of scenario development issues on operation page 300

Scenario development traces the progression of the ITER Research Plan. It is important that at each stage, the new scenarios that need to be explored can make use of the development and understanding gained in previous ITER operational phases. PFPO-1 will dedicate a significant fraction of experimental time to the assessment and optimization of control schemes that will be the basis of operation in PFPO-2 and later phases. For example, plasma initiation and start-up in the hydrogen phase will address several fundamental issues for all future operating scenarios and control schemes. Plasma breakdown, possibly assisted with ECRH, and early divertor formation will be tested. Model validation to enhance the confidence in predictive simulation of the next operational phase scenarios will accompany the execution of the ITER Research Plan.

The development of the ITER Research Plan will also benefit from validation of plasma scenario modelling tools on present tokamaks, including models for plasma initiation, control schemes, and models to describe the evolution of the temperature, density and current profiles. In particular, radial transport models for electron and ion energy and particles in Ohmic and L-mode plasmas that
were developed many years ago should be re-examined using the more extensive diagnostic capabilities of present devices so that they can be used to plan more optimal ITER ramp-up scenarios that maintain the internal inductance, li , within an acceptable range for control.

The basis for establishing ITER baseline Q = 10 DT operation, the initial major goal of the ITER Research Plan, is reasonably well established. But the experimental development of advanced (i.e. hybrid and steady-state) scenarios is more challenging, as they require operation at the limits of the hardware capabilities of any given tokamak where they are explored (i.e. operation at high power for a maximum time duration, close to plasma stability limits, etc.). With regards to operation of
ITER in regimes with enhanced core confinement, such as hybrid scenarios, but especially those that feature ITBs, none of the predictive models for such regimes are as yet in a position to make reliable projections. For the global scaling approach, the limitation may be intrinsic, in that the development and sustainment of ITBs depends on local plasma parameters (i.e. on detailed plasma profiles), which are not captured in scalar databases. For the transport models, while progress has been made in replicating ITB formation and sustainment, further work is required before projections can be made with confidence of such regimes to ITER.

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