DT Plasma Research Articles

A neutron camera for burning plasmas

Profile diagnosis of tokamak plasmas with neutron cameras is discussed, especially with regard to neutron detection limitations and possibilities in burning DT plasmas. A new detection method based on the magnetic proton recoil spectrometer technique is presented.

Neutron and plasma irradiations of fusion reactor materials using fusion plasma neutron sources

A compact and steady-state DT plasma with parameters relevant to a fusion-reactor, the FEF (fusion engineering facility), has been considered as a neutron source for the development of fusion reactor materials providing both. (1) 14 MeV fusion neutron irradiation testing, and (2) high charged particle flux for testing of plasma-facing materials during neutron irradiation. The 14 MeV neutron wall loading is about 1–2 MW/m 2, with a test section area of 1 m 2. One version of the test section is near the edge of the plasma in the vacuum chamber. Another is for high heat flux testing of plasma facing materials, where the charged particle flux leads to heat loads of several 10 kW/cm 2. New versions of FEF, with localized high neutron fluxes, are also proposed by introducing a sloshing ion distribution into the plasma.

On the possibility of neutron spectrometry for determination of fuel ion densities in DT plasmas

The fuel density (given by nd+nt and nd/nt) is one of the important parameters to determine and control in burning DT plasmas. Here, we propose that nd and nt be deduced from data on the dd and dt neutron energy distributions to be recorded with a neutron spectrometer having energy selective response to the plasma neutron flux. The potential diagnostic capabilities of the magnetic proton recoil spectrometer system are tested for the measuring conditions to be presented by the next step tokamaks such as ITER.

ZTI: Preliminary Characterization of an Ignition Class Reversed-Field Pinch†

A preliminary cost-optimized conceptual design of an intermediate-step, ignition-class RFP device (ZTI) for the study of alpha-particle physics in a DT plasma is reported. The ZTI design reflects p...

Stimulation of deuterium combustion in a DT plasma by secondary particles

Stimulation of deuterium combustion in a DT plasma by secondary particles

Extrapolation of tokamak energy confinement to Next Step Devices

The extrapolation of energy confinement to the conditions of a Next Step tokamak, to operate with an ignited DT plasma, is an essential ingredient of the data base needed for such a device. The present knowledge on energy confinement and the approaches used for providing a basis for the extrapolation are concisely described. The lines of further research are indicated which have to be followed to be able, in a few years, to extrapolate energy confinement to a Next Step tokamak with improved accuracy.

Net safety analyses and the European safety and environmental programme

The introduction outlines the Safety and Environment (S+E) part of the European Fusion Technology Programme and its relation to the NET device. The safety of NET proper is presented in terms of energy and radioactivity inventories, accident analyses, dispersion of radioactivity in the environment, radioactive wastes, and implications of component handling. The European S+E Programme is reflected by the expansion on these topics, as to a large extent, the basic information and quantitative results used have emerged from the collaboration between the European laboratories and NET under this programme. Where appropriate results acquired by the worldwide fusion community have been included to widen the understanding of the S+E aspects to be observed in the design of next generation fusion devices with burning DT plasmas.

Upgrading the JET Magnet System for 7MA Plasma

JET was designed for a plasma current of 5 MA and has operated successfully at that level. To enable JET to produce meaningful DT plasmas, it is necessary to upgrade the machine performance. The paper describes the effects on the poloidal and toroidal magnet systems of increasing the plasma current to 7 MA. It has not been necessary to increase the toroidal field but operation at higher plasma current increases the torque loading on the coils. In the case of the poloidal coils an increased flux swing is required so the magnetising current has been increased by 50%. Effects considered include magnetic forces and mechanical and thermal stresses in the coil. Modifications to the coil system and improvements to the power supplies that enable the new performance to be achieved are described. It is concluded that a 7 MA plasma current is feasible.

Design study of fusion experimental reactor at JAERI

Conceptual design studies of a next generation tokamak fusion reactor, the Fusion Experimental Reactor (FER) at JAERI are described. Recent design efforts have been focused on elaborating two typical design concepts, one based on the best existing physics data bases and the other on rather conservative physics basis, to be capable of achieving a self-ignited, long burning DT plasma and demonstrating the feasibility of key fusion reactor technologies. Various flexibility scenarios to enhance the machine capabilities to realize higher reactor-core performance have also been developed and incorporated in the design in accordance with an overall plan of phased construction and operation program of FER.

Another view of the control of the isotopic composition of initial DT plasmas in tokamaks

The present paper reviews phenomena associated with the implantation of hydrogen isotopes into graphite and carbonized layers, with particular emphasis on the implanted layer, i.e. the depth range where the injected atoms are deposited. Different modifications of graphite and carbonaceous layers are characterized with respect to their structure, and the influence of radiation damage is treated. Recent evidence is demonstrated for the occurrence of local hydrogen and methane molecule formation, and the diffusion of these molecules through the bulk.The ‘saturated layer’ formed during hydrogen implantation is modelled by a dynamic balance between ion deposition and trapping, thermal or ion-induced detrapping, and local recombination and molecular diffusion. Corresponding calculations yield satisfactory fits to monoenergetic implantation and thermal effusion experiments.The inventory in the saturated layer is only a small contribution to the total hydrogen inventory of carbon walls in fusion machines. Some recycling phenomena can be explained by the physics of the implanted layer, whereas others, like wall pumping, require different mechanisms for their explanation.

Respond to the article: “Another view of the control of the isotopic composition of initial DT plasmas in tokamaks”

Respond to the article: “Another view of the control of the isotopic composition of initial DT plasmas in tokamaks”

Part I Conceptual design of the R-tokamak

For the study of the alpha particles behavior in DT plasma the ‘Reacting Plasma Project’ (R-project in short) was initiated in 1981. The design study of the ‘R-tokamak’ was performed from 1981 to 1985, and evolved through three versions. The first version is based upon the usual tokamak and is similar to TFTR. The remote maintenance and dismantling however is considered very difficult. For the reduction of the residual radiation level the design of DT tokamak with aluminum alloy was conducted. It was found that hand-on maintenance after 1000 DT shots was feasible. Another effect of the aluminum alloy is the increase of the shell effect and stabilization of the variously shaped tokamak. The 3rd version tokamak with the extremely shaped cross-section (the crescent tokamak) is proposed for tokamak improvement (higher beta, better confinement and low radioactivity).

A method to control isotopic composition of initial DT plasmas in tokamaks

A method to control isotopic composition of initial DT plasmas in tokamaks

Part I. Conceptual design of the R-tokamak

For the study of the alpha particles behavior in DT plasma the ‘Reacting Plasma Project’ (R-project in short) was initiated in 1981. The design study of the ‘R-tokamak’ was performed from 1981 to 1985, and evolved through three versions. The first version is based upon the usual tokamak and is similar to TFTR. The remote maintenance and dismantling however is considered very difficult. For the reduction of the residual radiation level the design of DT tokamak with aluminum alloy was conducted. It was found that hand-on maintenance after 1000 DT shots was feasible. Another effect of the aluminum alloy is the increase of the shell effect and stabilization of the variously shaped tokamak. The 3rd version tokamak with the extremely shaped cross-section (the crescent tokamak) is proposed for tokamak improvement (higher beta, better confinement and low radioactivity).

ICRF fusion reactivity enhancement in tokamaks

A systematic study of ICRF fusion reactivity enhancement has been conducted, using a new bounce-averaged two-dimensional Fokker-Planck code. Second-harmonic heating of deuterium in a 50–50 DT plasma is assumed, and the results are obtained as a function of background plasma density and temperature. An enhancement factor of ten is achieved at low Q (= fusion power/RF power), which is important for ion-tail diagnostics, but at Q = 0.5 the enhancement is ⪅ 2. Significant poloidal variations in ion density (up to 14%) and in fusion reactivity (by a factor up to 2.5) are found.

Shielding Analysis for the Horizontal X-ray Imaging System on the TFTR

Intense neutron and gamma radiation will accompany the reactor like DD and DT plasmas that are planned to be generated on the TFTR. Since many plasma diagnostic systems use detectors that are sensitive to neutrons and gamma-rays to various degrees, local shielding will be required to achieve acceptable signal to noise ratios during those high power operations. In this paper, we analyze the shielding problems related to the horizontal x-ray imaging system.

Status of Fusion Experimental Reactor (FER) Design

Conceptual design studies of the Fusion Experimental Reactor (FER) have been conducted at JAERI in line with a long-range plan for fusion reactor development laid out in the long-term program of the Atomic Energy Commission issued in 1982. The FER succeeding the tokamak device JT-60 is a tokamak reactor with a major mission of realizing a self-ignited long-burning DT plasma and demonstrating engineering feasibility. The paper describes recent developments of the FER design concept.

Fusion Product Diagnostics in Magnetic Confinement System

Fundamentals and present status of fusion product diagnostics in the case of magnetic confinement system are reviewed. Methods of diagnosing fusion products, i. e. neutrons and fast charged particles, are presented for DD, D3He and DT plasmas.

A DT-burning upgrade to MFTF-B

The MFTF-B Upgrade will provide a conclusive demonstration of plasma confinement in the advanced tandem mirror geometry proposed for a fusion reactor and will also provide the capability for operating a power-producing section of a fusion reactor for periods up to 100 h with neutron wall loadings of 1–2 MW/m2. The necessary physics data base for reactor design will be obtained during initial operation at Q in the range of one to two in DT plasma at which point about one-third of the center cell heating is provided by the alpha particles. In the latter operation, the MFTF-B program will shift from resolution of physics issues to integration of all subsystems in a nuclear environment and development of reactor blankets.

On the ignition of a self-sustained fusion reaction in a dense DT plasma

Ignition of a self-sustained fusion reaction in a strong magnetic field is studied. The critical ignition dimensions and the threshold ignition energy are calculated as functions of the fuel density and the magnitude of the transverse magnetic field. A new method for producing ultra-high magnetic fields is proposed which provides at once heating of the fuel by induction currents and localization of the micro-explosion by the magnetic field. Simple analytic estimates show that a certain combination of inertial, magnetic and wall methods of plasma confinement may decrease the ignition energy below 100 kJ.