Fusion Power Production Research Articles

Multi-Mode transport modeling of the International Thermonuclear Experimental Reactor (ITER)

Predictions are made for the performance of the International Thermonuclear Experimental Reactor (ITER) [R. Aymar, V. Chuyanov, M. Huguet, R. Parker, and Y. Shimomura, in Proceedings of the Sixteenth International Atomic Energy Agency Fusion Energy Conference, Montréal, Canada 1996 (International Atomic Energy Agency, Vienna, 1997), Paper IAEA-CN-64/01-1] design using the Multi-Mode model in the time-dependent one- and one-half-dimensional (1-1/2-D) BALDUR [C. E. Singer et al., Comput. Phys. Commun. 49, 275 (1988)] transport code. This model predicts the temperature and density profiles observed in present-day tokamak experiments more closely on the average than other models currently available. Simulations using the Multi-Mode transport model, with its inherent gyro-Bohm scaling, indicate that ITER will ignite, even with edge temperatures as low as 0.25 keV (L-mode, or low-confinement mode, boundary conditions) or with volume averaged density as low as 0.775×1020 m−3 (just below the Greenwald density limit, when Tedge=0.75 keV). The ignition is found to be thermally stable, and the fusion power production is easily controlled by varying plasma density, impurity content, or edge temperatures. The nonequilibrium impurity radiation model used in these simulations predicts that a significant fraction of the fusion power is radiated when conditions are close to marginal ignition.

TFTR DT experiments

The Tokamak Fusion Test Reactor (TFTR) is a large tokamak which has performed experiments with 50:50 deuterium - tritium fuelled plasmas. Since 1993, TFTR has produced about 1090 D - T plasmas using about 100 grams of tritium and producing about 1.6 GJ of D - T fusion energy. These plasmas have significant populations of 3.5 MeV alphas (the charged D - T fusion product). TFTR research has focused on alpha particle confinement, alpha driven modes, and alpha heating studies. Maximum D - T fusion power production has aided these studies, requiring simultaneously operation at high input heating power and large energy confinement time (to produce the highest temperature and density), while maintaining low impurity content. The principal limitation to the TFTR fusion power production was the disruptive stability limit. Secondary limitations were the confinement time, and limiter power handling capability.

H-mode confinement and fusion performance in JET

JET has conducted extensive experiments in H-mode plasmas focusing on two broad aims. The first, involving the characterization of reactor-relevant aspects of the regime, encompasses the study of steady-state H-modes at plasma currents of up to 6 MA. In these plasmas, energy confinement is typically a factor of 1.8 - 2 greater than predicted by L-mode scalings and local transport varies with the ion Larmor radius, at least for input powers well above the H-mode threshold power. Using nitrogen impurity seeding, a regime has been developed in which good core confinement, with an enhancement of 1.5 relative to L-mode scaling, is combined with radiated power fractions of up to 80%. The second major goal is to optimize fusion power production by exploiting high-confinement ELM-free H-modes, which exhibit energy confinement enhancement factors of three or more relative to L-mode plasmas. In this regime, fusion power quotients have been achieved, though the highest fusion performance is transient.

Effect of the neutral beam fuelling profile on fusion power, confinement and stability in TFTR

The stored energy and neutron yields of TFTR plasmas have been modelled, with small dispersion, by considering a parameter associated with the peakedness of the beam fuelling profile as an independent variable. Scalings with fixed multiplicative constants and exponential powers determined from the experimental data are applied to deuterium discharges at different major radii in TFTR, and also to deuterium-tritium (DT) plasmas. The domain of the data considered includes the supershot, high poloidal beta, L mode and limiter H mode operational regimes, as well as discharges with a reversed shear (RS) magnetic configuration. The role of plasma parameters, such as plasma current Ip, edge safety factor qa and toroidal field Bt, are considered in this study. Improved energy confinement and neutron production are relatively insensitive to these parameters compared with the beam fuelling peakedness parameter when considering plasmas that are stable to magnetohydrodynamic modes. In contrast, plasma stability is dependent on these quantities and also on parameters describing the peakedness of the equilibrium profiles. On the basis of externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path towards a discharge condition with fusion power gain Q>1 is suggested

The stability of advanced operational regimes on the Tokamak Fusion Test Reactor

The performance of the Tokamak Fusion Test Reactor [D. Meade and the TFTR Group, in Plasma Physics and Controlled Nuclear Fusion Research, Washington, D.C., 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. I, pp. 9–24], as defined by the maximum fusion power production, has been limited, not by confinement, but by stability to pressure-driven modes. Two classes of current profile modification have been investigated to overcome this limit. A new technique has been developed to increase the internal inductance of low-q (q≈4), high-current (Ip>2MA) plasmas. As was the case at higher edge q, the disruptive β limit has been found to increase roughly linearly with the internal inductance, li. Plasmas with hollow current profiles, i.e., reversed shear, are also predicted and experimentally observed to have increased stability in the negative shear region to ballooning and kink modes. However, performance of these plasmas is still limited by pressure-driven modes in the normal shear region.

Neutron activation for ITER

There are three primary goals for the Neutron Activation system for ITER: to maintain a robust relative measure of fusion power with stability and wide dynamic range (seven orders of magnitude), allow an absolute calibration of fusion power production, and provide a flexible and reliable system for materials testing. The nature of the activation technique is such that stability and wide dynamic range can be intrinsic properties of the system. It has also been the technique that demonstrated (on JET and TFTR) the most accurate neutron measurements in DT operation. Since the detectors for assaying the radioactivity are not located on the tokamak and are therefore amenable to accurate characterization, and if the activation samples are placed very close to the ITER plasma with minimal scattering or attenuation, high overall accuracy in the fusion energy production (7%–10%) should be achievable on ITER. In the paper, a conceptual design is presented. A system is shown to be capable of meeting these three goals, and unresolved design issues are identified.

Disruption Avoidance on TFTR

Disruptions on the Tokamak Fusion Test Reactor (TFTR), especially those occurring at high stored energy, result in lost experimental run time because many discharges are required to regain wall conditions necessary for good plasma performance. A variety of disruption types have been observed on TFTR. These include density-limit disruptions, those caused by a high influx of impurities, those occurring during the current rampdown, those resulting from locked modes, and those occurring at high normalized β (β N = β T αβ T /I p ). A combination of operational experience and limiter development has helped to avoid many potential disruptions. However, the experimental goal of high fusion power production engenders the risk of high-β N disruptions. A system to limit β N by reducing the neutral beam power as a preprogrammed β N limit is reached is now in use to help avoid high-β N disruptions. Operational issues of disruption avoidance, the β N feedback system, the limitations and possible improvements of the system are discussed.

Electroslag remelt processing of irradiated vanadium alloys

This paper describes experimental efforts to investigate the potential of a slag remelting process for reducing radioactivity of irradiated vanadium alloys used in a fusion power production facility. The experiment determined the removal characteristics of four surrogate transmutation isotopes significant to accident safety expected in a V5Ti5Cr alloy irradiated under fusion conditions (Ca, Y, to simulate Sc, Mn, and Ar). Removal of these isotopes could decrease the accident risk of reprocessing irradiated vanadium and reusing it in fusion reactors. An electroslag remelt furnace was used in the experiment to melt and react the constituents using a calcium fluoride slag. The process achieved 90% removal of calcium and over 99% removal of yttrium. Analyses indicate that 40% of the manganese has been removed. Argon analysis of the refined ingots indicates that 99% of the argon was removed from the vanadium alloy.

Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium–tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 1021 m−3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high-performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral-beam heating is begun.

Dynamic ignition - a new operation criterion for fusion plasmas

Burn dynamic studies of magnetically confined fusion plasmas have shown that - for certain fueling regimes - there exists a stable equilibrium point in the plasma density-temperature state plane associated with high fusion power production. In approaching this operation point the plasma is heated by auxiliary mechanisms. The question of interest is: When can the external heating power be turned off so that the plasma temperature and density will evolve into the stable equilibrium point only due to internal fusion power deposition? The associated system dynamics suggest the definition of a new ignition criterion which seems to be more adequate for accomplishing steady-state operation of fusion reactors than todays ignition criteria.

Recent D-T results on TFTR

Routine tritium operation in TFTR has permitted investigations of alpha particle physics in parameter ranges resembling those of a reactor core. ICRF wave physics in a DT plasma and the influence of isotopic mass on supershot confinement have also been studied. Continued progress has been made in optimizing fusion power production in TFTR, using extended machine capability and Li wall conditioning. Performance is currently limited by MHD stability. A new reversed magnetic shear regime is being investigated with reduced core transport and a higher predicted stability limit.

Using the shield for thermal energy storage in pulsar

The PULSAR pulsed tokamak power plant design utilizes the outboard shield for thermal energy storage to maintain full 1000 MW(e) output during the dwell period of 200 s. Thermal energy resulting from direct nuclear heating is accumulated in the shield during the 7200 s fusion power production phase. The maximum shield temperature may be much higher than that for the blanket because radiation damage is significantly reduced. During the dwell period, thermal power discharged from the shield and coolant temperature are simultaneously regulated by controlling the coolant mass flow rate at the shield inlet. This is facilitated by throttled coolant bypass. Design concepts using helium and lithium coolant have been developed. Two-dimensional time-dependent thermal hydraulic calculations were performed to confirm performance capabilities required of the design concepts. The results indicate that the system design and performance can accommodate uncertainties in material limits or the length of the dwell period.

Fusion-A Potential Power Source

Duplicating the fusion process of the sun and the stars for energy production on earth would present many difficulties. The state of matter at such temperatures--the plasma state--may be considered a gas of electrons and nuclei, so one problem is the need to confine a hot, reacting plasma. Because the plasma is an electric conductor, it is subject to magnetic forces. Thus, one approach is to confine the hot plasma by a magnetic field. Another approach is to heat the matter so rapidly that the fusion reactions take place before the matter has had time to fly apart, that is, to use inertial confinement. At the United Nations' Atoms for Peace Conference in 1958, a remarkably cooperative, international research effort began. In spite of many difficulties, substantial progress has been made. Initially, many tokamaks were built with circular cross sections. However, shaped plasmas were shown to have clear advantages. The cross sections of some of the larger ones are illustrated here. The two largest devices in the US are the Tokamak Fusion Test Reactor (TFTR) at Princeton and the Doublet III-D (DIII-D) at General Atomics in San Diego. The TFTR device is constructed with neutron shielding and equipped to handlemore » the superheavy hydrogen isotope tritium, which is radioactive. This makes it possible to operate the device with the optimum fuel mixture: an equal mixture of deuterium and tritium. This mixture is optimal because the cross section for the DT reaction has by far the largest cross section of the fusion reactions mentioned above. A large effort is presently under way to design the International Thermonuclear Experimental Reactor (ITER). This is a joint effort by the European Community, Japan, Russia, and the US. Goals include the production of fusion power in excess of 1,000 MW for studying the physics of igniting plasmas, and the integrated demonstration of fusion-reactor technologies.« less

Fusion power production from TFTR plasmas fueled with deuterium and tritium.

Peak fusion power production of 6.2±0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2×1017 m−3 without the appearance of either disruptive magnetohydrodynamics events or detectable changes in Alfvén wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits.Received 10 February 1994DOI:https://doi.org/10.1103/PhysRevLett.72.3526©1994 American Physical Society

Reduced activation ferritic alloys for fusion

Abstract Reduced activation Martensitic alloys can now be developed with properties similar to commercial counterparts, and oxide dispersion strengthened alloys are under consideration. However, low chromium Bainitic alloys with vanadium additions undergo severe irradiation hardening at low irradiation temperatures and excessive softening at high temperatures, resulting in a very restricted application window. Manganese additions result in excessive embrittlement, as demonstrated by post-irradiation Charpy impact testing. The best composition range for Martensitic alloys appears to be at 7 to 9 Cr and 2 W, with swelling of minor concern and low temperature irradiation embrittlement perhaps eliminated. Therefore, Martensitic reduced activation steels in the 7 to 9 Cr range should be considered leading contenders for structural materials applications in power-producing fusion machines.

Latest results from JET

Plasma-wall interactions have been important in many aspects of JET's programme during the past year. A major achievement has been the first production of significant fusion power from the deuterium-tritium reaction. Techniques of tritium handling and recovery have been developed and tested. The tokamak has operated extensively in the X-point configuration and comparisons have been made between carbon and beryllium target materials. Experiments have been made to ameliorate the heat load on the divertor targets by strong gas puffing as well as by careful tile design and alignment. ELMs have been utilized to produce a quasi steady-state H-mode with duration of about 20 s.

Pilot plant: An affordable step toward fusion power

Progress is reported on a study to define a “pilot plant” to demonstrate the production of high grade heat in a fusion power plant configuration at the lowest possible capital cost. We are considering several driven reactor tokamak designs with fusion power production levels in the 15–50 MWth range, using demountable copper coils. We conclude that it is acceptable for such facilities to be net consumers of electricity as a trade-off to achieve low capital cost, which we estimate to be in the $1 billion range. These designs are based on currently accepted physics models. Even lower cost designs may be possible, if we depart somewhat from the current physics database.

I. Burning Plasma Experiment Physics Design Description

The mission of the Burning Plasma Experiment (BPX) is to determine the physics behavior of self-heated fusion plasmas and demonstrate the production of substantial amounts of fusion power. Supporting objectives, included in this paper are: demonstrate the production of fusion power in excess if 100 MW, a fusion-reactor-level power density; determine the confinement physics, operational limits, and alpha-particle dynamics of self-heated fusion plasmas with alpha power greater than auxiliary heating power; demonstrate heating, fueling and plasma handling techniques necessary to produce reactor-level power density, self-heated fusion plasma; optimize plasma performance in the range of Q = 5 to ignition, with fusion power up to 500 MW (here Q is defined as the ratio of fusion power to externally supplied plasma heating power, including ohmic heating power).

System Studies of Copper- and Superconducting-Coil Pilot Plants

The TETRA systems code is used to examine devices with both normal copper and superconducting coils as vehicles for steady-state production of fusion power in a Pilot Plant. If the constraints of p...

Normal Flow Heat Exchanger for Divertor Panel Cooling

Helium cooling offers safety advantages for fusion reactors because helium is single-phase and chemically inert. The Normal Flow Heat Exchanger (NFHX) enables highly efficient heat transfer to helium with low pressure drops at high heat fluxes. Thus, the NFHX enables helium cooling of divertor panels in Tokamak reactors and potentially enhances the safety of fusion power production. This paper explains the NFHX concept, presents the current status of the technology, and shows how the NFHX can be used to cool divertor panels in Tokamak reactors.