H. Knoepfl

On Mechanical Boundary Conditions in Large Helias Coil Systems and the Influence of Sliding on Stresses

The WENDELSTEIN 7-X stellarator experiment is developed as a Helias (Heli cal A dvanced S tellarator) configuration. The superconducting modular coils with nonplanar shape introduce a magnetic field with a maximum net coil force of about 4 MN. The resultant force vector of a whole field period is directed towards the torus centre. In order to support the magnetic forces, a scheme of mutual support has been developed. Each coil is enclosed by an individually adapted stainless steel housing. The coils of a field period are connected to a module by a pair of reinforcements inside the torus. For reasons of accuracy calculations are made within one period. The investigated stress levels in the coils and support depend on the mechanical contact between winding pack and coil housing. The nonplanar coils lead to difficulty in defining the bearing conditions at the ends of a period. The assumptions for prescribing these boundary conditions, which are needed for the FE analysis, are presented.

Tritium Inventories Used for NET Safety Analysis

The NET Predesign Phase and the ITER Conceptual Design Activity resulted in documented conceptual designs. Assessment of these reports led to the quantification of a reasonable range of values for the tritium inventories of the following systems or plant areas: fuelling, plasma vacuum pumping, fuel purification, blanket tritium recovery, fuel cycle auxiliaries (including isotopic separation system), first wall and divertor, dust produced inside the plasma chamber, blanket, and coolant water of first wall, divertor and blanket. For realistic dose calculations source terms are derived from inventory estimates. Release limitation by process design features and confinement strategy following specific accident scenarios have been considered. The potential to reduce source terms by design improvements and alternative confinement strategies are evaluated.

EFFECTS OF THE EDDY CURRENTS FOR SOME DESIGN AND INTERPRETATION PROBLEMS IN A TOKAMAK

The P.F. coil current and voltage scenario in a tokamak device is generally designed and preprogrammed disregarding eddy currents, which are considered as disturbances to be taken into account in a second stage, i.e. when making the analysis of the expected performance of the machine. In addition, eddy currents also affect the magnetic measurements used for the plasma shape identification, which is generally performed with the aid of static MHD equilibrium codes. The aim of the present paper is then twofold: - show how eddy currents can be entered in a procedure for the definition of the plasma scenario; - show how a plasma identification procedure can include the presence of not measurable eddy currents. The assessment of the advantages of this new approach is carried out by comparing some results obtained with and without the presence of eddy currents; ASDEX-U geometry has been considered for the examples, because of its relevance for next step ITER-like tokamaks.

RUNAWAY ELECTRONS ON PLASMA FACING COMPONENTS

Runaway electrons can cause severe damage to plasma facing components of large tokamaks. The designs proposed for the first wall and divertor of the next large fusion experiment, ITER (International Thermonuclear Experimental Reactor), are investigated. Energies of up to 300 MeV per electron and surface energy depositions of 30 MJ/m2 are assumed. The GEANT code originating from high energy physics was used to model the energy deposition [J/cm3] quantitatively as a function of the penetration depth and material. The magnetic field was included in the analysis. The energy deposition in the bulk material for a given surface energy load is roughly independent of the incident angle and energy (above 100 MeV) since the main physical process of the energy loss is the formation of an electromagnetic shower, i.e. rapid dissipation of the initial energy into many electrons, positrons and photons. Typical divertor designs protect the cooling tubes with a 1 cm thick graphite layer. Melting of such molybdenum (copper) cooling tubes occurs at a heat load of 50 (25) MJ/m2. Every additional cm of graphite roughly doubles the runaway protection. Since it is proposed to operate ITER with low cooling water temperatures (TH2O ≈ 100°C), water pressurization due to runaway electron impact is not a serious problem if the cooling pipes do not melt. If the first material facing the plasma is metallic, melting must be expected for heat loads of around 15 MJ/m2.

CONCEPTUAL AND EXPERIMENTAL STUDIES OF ELECTRICAL JOINTS FOR NET-ITER COILS

ABSTRACT The interpancake joints of superconducting cables for the Central Solenoid of NET-ITER are special components for which contradictory requirements have to be fulfilled, regarding electrical mechanical and thermal properties. A particular attention has to be paid on the stability of the conductor in the joint during a plasma disruption because of the large magnetic field variation at the joints location(> 2T within about 0.1 s). From this point of view, the joint design appears as a challenge when a low dc resistance and a low available space have to be taken into account. Two NET-CEA contracts have been launched in order to develop and test subsize Nb3Sn joints under representative conditions. The first set of joints has been dedicated to basic parametric studies from which general properties of different kinds of joints can be drawn. Two main designs have been considered: the so-called butt-joint and overlap (subdivided) joint. Three different samples of each kind are tested under bath cooled conditions for checking dc resistance, dc losses and behaviour under varying magnetic field. This experimental programme has been associated to a theoretical analysis necessary to extrapolate the results to full size joints.

RADIOACTIVITY CONFINEMENT FOR LOSS OF COOLANT ACCIDENTS IN NET

This paper provides an overview of confinement pressurization for the case of Loss of Coolant Accidents (LOCAs) occurring during plasma burn as well as during conditioning and baking. The radioactivity confinement strategy is outlined.

DESIGN AND SAFETY CRITERIA FOR REACTOR RELEVANT HEAT TRANSPORT SYSTEMS

The candidate media (helium and water) for the primary cooling systems of a fusion plant are compared and criteria for the selection of the appropriate coolant developed. Robust formulas are developed to determine the space requirements for the cooling system. The formulas expressing the volumes of a cooling loop are derived from correlations for heat transfer, pressure drop, mass and energy balances. They define the required coolant volumes as a function of characteristic design parameters such as thermal power, temperature increase of the coolant, the temperature difference across the heat exchange surfaces and pumping power. The formulas are validated by comparison with design values of helium cooled fission reactors (High Temperature Reactor HTR) and a pressurized water cooled fission reactor (PWR). For the case of pipe break, the expansion volumes required to avoid an excessive overpressure in the building are calculated (important safety criterion). Mixing of the coolant with air contained in the expansion room is assumed. If high operation temperatures could be realized for the helium cooling (THe, max > 450°C) this would have a safety advantage compared with water cooling at conditions typical of a PWR. For lower allowable inlet temperatures, water requires a smaller expansion volume than helium, at the expense of energy conversion efficiency.

Controllability of NET Plasmas After Loss of Beta Events

The aim of this work is to study the controllability of NET double null plasmas with respect to vertical displacements. The controller used has been designed for a reference ignited plasma. As the plasma poloidal beta drops due to soft or hard disruptions the plasma is likely to become more unstable and controller output voltage might exceed the amplifier linear range. We found that for saturation of 1 kV the controllability range as parametrized by allowable initial displacements exceeds 10 cm if the plasma pressure does not decrease. For low beta plasmas the controllable disturbances are smaller and the system is uncontrollable if the plasma current drops together with the pressure.

PASSIVE REMOVAL OF AFTERHEAT IN THE NEXT STEP FUSION DEVICE

In the next-step fusion device (e. g. ITER) a 14 MeV wall load of about 1 MW/m2 will cause strong activation of the plasma chamber. This involves heat generation by radioactive decay of the resulting nuclides. In the absence of cooling this afterheat leads to a moderate temperature increase with possible consequences to chemical release rates. Therefore a realistic prediction of the expected maximum temperatures by a permanent and complete loss of all active cooling is important. Two distinct bidimensional models, one midplane (r,φ) and the other axial (r,z), were developed at JRC Ispra and IPP Garching respectively. Both models consider the ITER-concept with shielding blanket and produce quite similar results. The temperatures always remain below 700°C. The tungsten divertor temperature dominates only after some hours and up to two days after the loss of cooling. Beside some parametrical studies, the effect of allowing natural convection of air in the magnet-cryostat-interspace is presented. A simple extrapolation to a fusion reactor is discussed.

Validation of Numerical Codes for the Analysis of Plasma Discharges in a Fusion Device

Electromagnetic aspects play a central role in the design of ITER-like reactors and call then for an extensive use of complex and advanced numerical codes; therefore the validation of the code predictions against experiments has to be considered as a strict requirement. Aim of the paper is to give a comprehensive presentation of this problem in the light of the results of a campaign of validation runs. The main outcome of this work is that a number of computational procedures, which have been developed for the NET project and then extensively used also for ITER studies, can be considered as experimentally validated in a sufficiently wide range of cases of interest. In particular, computed values of magnetic signals are compared with experimental measurements made during some typical ASDEX-U discharges. From the electromagnetic point of view, many features of this machine are common to the ITER concept, so that the results of the validation can reasonably be extended to the ITER case.