[en] The TBM structural material is a low activation martensitic steel relevant for the future DEMO reactor development. This is a ferromagnetic steel with high saturation magnetic flux density. Therefore two categories of electromagnetic forces are expected on the TBM during all the ITER operating scenarios; the Lorentz Forces (LF) and the Maxwell Forces (MF) that apply to a magnetizated body. The MF can be caused by two reasons: a misalignment of the body magnetization with the external field originating a torque and by a magnetic field not uniform in space originating net resultant forces directed toward the increasing external field. To evaluate the effect of these loads on the TBM structure and attachment system, detailed electromagnetic analysis have been carried out and reported in this paper. Using ANSYS code several 3D models have been developed for both static and transient analysis. According to the ITER loads specification document a plasma current disruption of type II (linear decay in 40 ms) has been identified as the most demanding event for the TBM equatorial position and its effects have been investigated. LF on TBM have been evaluated and eddy currents distributions in the surrounding components have been stored for each time step. Using these currents as external loads in static analysis, the evolution of Maxwell forces during the plasma disruption have been investigated. While the maximum of resultant forces is reached at the End Of Burning time, just before the disruption beginning, the maximum of torque is obtained at the end of the disruption when the misalignment between the TBM magnetization (supposed to be rigid and thus anchored to the field direction at the disruption beginning) and the external field is maximum. At the end the effect of the toroidal field ripple on the MF has been evaluated taking into account the real geometric shape and discreteness of toroidal field coils. In spite of the small field perturbation produced by the ripple, the strong space gradient of this perturbation produces a very significant effect (an increase of about 50%) in the net resultant Maxwell force on the TBM. (orig.)

[en] In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group, the Helium Cooled Pebble Bed Test Blanket Module- (HCPB-TBM) System is developed. The TBM test schedule foresees four different campaigns for simulation of DEMO relevant conditions, campaign requires a dedicate TBM. Therefore a concept for TBM integration into ITER is designed with attention to simplify the mounting/dismounting operations. This paper presents the status of this concept with regard to the operations in hot cell required to install a new TBM into an equatorial TBM Port Plug (PP). This includes the establishment of the connection for the attachment, supply- and diagnostic lines in the environment of the interface (IF 1) between the TBM rear part and the PP backside shield. The connection of IF 1 has to be designed to cope with a temperature difference between TBM and PP (∝200 K) and the EM-loads during normal operation and disruption scenarios. The reference attachment concept based on shear keys and flexible cartridges is revised to cope with new conditions on the load and at the interface to the PP. According to the latest results of EM analysis, a radial component of the Maxwell forces (due to the ferromagnetic structural material) has been identified as an additional challenging load for the attachment. Furthermore, the replacing operations at IF 1 are influenced by the design of the PP; the recent ITER proposal based on a removable back side shield allows access to the IF 1 from the periphery after the frame of the PP surrounding the TBM is removed. As for the mechanical attachment, the tools and operations for connection of the TBM supply lines (Helium-, Purge- and measurement lines for different purpose depending on the test schedule) are strongly influenced by the restrictions to access IF 1, too. Dismantling of the frame would allow direct access to the interface by e.g. orbital welding tools. The concept for connection of the TBM diagnostic lines does not foresee an interface between the TBM and the PP back side shield because of the very restricted space conditions. Therefore the diagnostic lines will be routed inside of a pipe which is attached to the TBM rear part. This instrumentation pipe is designed to penetrate the whole radiation shield up to the interface between the PP back side shield rear part and the Ancillary Equipment Unit (AEU). At this interface the diagnostic lines exit the instrumentation pipe by a feed through where they are connected to a multi plug which provides the connection to the Data Acquisition System. The vacuum boundary between the back side shield and the instrumentation pipe will be provided by a bellow. After a consistent concept for the integration of the HCPB TBM in ITER has been developed, further investigation will be needed to develop tools and procedures which are required to install the TBM into the PP during the maintenance and refurbishment operations in the hot cell. (orig.)

[en] Full text: The design of the ITER Blanket System is controlled by three main considerations. Two of these considerations, plasma heat flux to the surface and nuclear heating, determine the cooling requirements and coolant distribution to control thermal stress in the modules. Electromagnetic forces due to off-normal events like disruptions must be controlled to be within the strength of the supports on the vacuum vessel and determine the mechanical stress in the blanket components. First, electromagnetic forces are generated by three different causes during an off-normal event. The shift in plasma paramagnetism during thermal quench generates eddy currents due to the change in toroidal field. Second the changing plasma current during current quench generates eddy currents due to the change in poloidal field. Finally, halo currents flow from the plasma to the wall due to plasma motion during current quench. The analyses conducted are based on disruption simulations , performed using the DINA code. The DINA output includes information on toroidal flux change during thermal quench, plasma shape, current, and position during current quench, and halo current flow to the plasma-facing surface. Raw DINA output was processed to reduce the plasma current temporal and spatial variation to time dependent current in a fixed set of 64 conductors that surround the plasma volume. The OPERA code was used to simulate the transient eddy currents in, and current flow through the complex 3D blanket components. Forces and moments were calculated from the current distribution and the local magnetic field. We studied numerous options for slits to control eddy currents in the shield blocks, electrical connections between the first wall and shield, fault scenarios for insulators in the support system between the blanket and the vessel, and variations on two different types of first wall panels (lower and higher heat flux rating). The support system between the first wall and shield block and between the shield block and vessel contains several insulators to control the flow of halo or eddy currents among the various components. We have examined the consequence of failure of one or more insulators by replacing the insulator with a controlled electrical contact. Insulator failure creates addition conducting loops or current paths. These studies led to revision of the design of the mounts. (author)

[en] The ITER Toroidal Field Coil (TFC) system is made of 18 D shaped coils spaced by 20 in toroidal angle, this discontinuity can cause significant losses in the confinement of high energy particles (a-particles or high-energy ions from neutral beam injectors) due to their trapping inside the ''ripple'' valleys; the toroidal ripple of the flux surfaces in contact with the First Wall (FW) produces unwanted peaking in the heat loads on the FW itself. Due to these reasons an accurate evaluation of the Toroidal Field Ripple (TFR) in various operation conditions has been performed. To this end, various Finite Element Models (FEM), using the ANSYS code, have been developed. To produce regular field mapping, these models make use only of structured meshes that allow high filled precision and a very regular spacing of the model elements. The mapping has been extended to all the region internal to the FW including the FW itself. The FEM takes into account the real 3-D shape of the TFC. The FEM model of TFC is made of three nested D shaped coils capable of reproducing with high accuracy the real geometry of the TFC. The value found of the TFR has confirmed the need of introducing some correcting elements. The benefit of introducing Fe inserts between the two vessel shells at the outboard has been tested. It has been shown that excluding the inserts from the equatorial region, as it was made in all the previous works on this subject, does not allow any significant benefit, it has instead proven that a ripple reduction up to a factor 3 or more could be obtained including this region. The poloidal field produces a misalignment between the magnetization of Fe insert, in stationary conditions aligned to the resultant field, and the toroidal field. The effect of this misalignment on the ripple correction produced by the inserts has been checked and was proven to be negligible. The possible ripple over-compensation during plasma reduced scenario (at halved toroidal field) has been analyzed. At the end, the field perturbation introduced by the presence of a Test Blanket Module (TBM) for DEMO in the equatorial port has been analyzed. It has been shown that the TMB (made of about 2.7 tons of EUFER and with saturated magnetization about 1.9 T) introduces a very large field perturbation: about three times the uncorrected and ten times the TFR corrected with the inserts. In order to allow the analyses of the particle losses and of the heat loads, a detailed ripple map of the TFR has been produced for the whole region inside the FW and for all the main cases that have been analyzed: a) without inserts and without TBM, b) with inserts and without TBM, c) without inserts and with TBM, d) with inserts and with TBM. The relative precision in the error field obtained in these analyses is better than 1%. (orig.)

[en] Static and transient analyses have been carried out to evaluate Electro Magnetic loads on the European Helium Cooled Pebble Bed Test Blanket Module (HCPB-TBM) during a type II linear disruption (Lorentz forces) and (because of the presence of ferromagnetic structural material) during normal ITER operating scenario (Maxwell forces (MFs)). At last the influence of the toroidal field ripple on the MFs has been investigated

[en] Full text: The causes of plasma asymmetries and rotation during disruptions are still an open issue even though their effects are clearly seen on present machines like JET where the vessel has been observed to move horizontally during asymmetric VDEs. Strong horizontal forces are then expected to be related to the plasma asymmetries. In ITER, loads caused by asymmetric VDEs are expected to be among the highest mechanical loads. A model consistent with most of JET measurements has been developed assuming that the asymmetric loads are caused not by a direct exchange of current between plasma and structure (as in the case of halo or surface currents) but to asymmetric conductive paths which arise, in the structures, when the plasma column asymmetrically wets the wall. This model of asymmetric toroidal eddy currents (ATEC) has been implemented in detailed finite element (FE) electromagnetic analyzes of locked and rotating AVDE experienced at JET. The results showed substantial match with all the main asymmetry related measurements done at JET. The same ATEC model is then used to assess loads on the ITER VV during asymmetric VDEs and detailed results are reported and discussed in this paper. (author)

[en] The integration system for installation of the European Helium Cooled Pebble Bed Test Blanket Module (EU HCPB TBM) in ITER being developed in the frame of the FZK Fusion programme uses three main interfaces to connect the TBM to its sub-systems. This paper describes how the occupation of the ITER hot cell for TBM installation into the port plug can be limited to the essential taking into account the update of the port plug design. The new design allows the dismantling of the radiation shield from the port plug frame. If a new shield is used for each new TBM the operations for connection of interface I between the TBM and shield can be done outside of the ITER hot cell as hands on operation

[en] In ITER two heating (HNBI) and one diagnostic neutral beam injectors (DNBI) are foreseen. Inside these components there are very stringent limits on the magnetic field (the flux density must be below some G along the ion path and below 20 G in the neutralizing regions). To achieve these performances in an environment with high stray field due to the plasma and the poloidal field coils (PFC), both passive and active shielding systems have been foreseen. The present design of the magnetic field reduction systems (MFRS) is made of seven active coils and of a box surrounding the NBI region, consisting of ferromagnetic plates. The electromagnetic analyses of the effectiveness of these shields have been performed by a 3D FEM model using ANSYS code for the HNBI. The ANSYS models of the ferromagnetic box and of the active coils are fully parametric, thus any size change of the ferromagnetic box and coils (linear dimension or thickness) preserving the overall box shape could be easily reproduced by simply changing some parameter in the model

[en] An accurate evaluation of the toroidal field ripple in ITER has been carried out by finite element models including the presence of the 18 TF coils and a set of ferromagnetic inserts that aim to lower the field ripple well below 1%. As shown, a set of ad hoc distributed plates made of AISI 430 stainless steel (B_sat = μ₀M_sat ∼1.5 T) and located at the outboard plasma region side in between the vessel shells can reduce the peak ripple at the plasma boundary to ∼0.4% at full toroidal field (i.e., B_TF ∼5.3 T at R = 6.2 m). Better compensation can be achieved by adopting higher magnetic saturation materials (e.g., EUROFER). The Test Blanket Modules pair modeled here and made of EUROFER (B_sat ∼1.8 T at 300 K) introduces a large perturbation to the field ripple up to ∼1.1% at full field

[en] Due to a number of causes (the finite number of toroidal field coils or the presence of concentrate blocks of magnetic materials, as the neutral beam shielding) the actual magnetic configuration in a Tokamak differs from the desired one. For example, a ripple is added to the ideal axisymmetric toroidal field, impacting the equilibrium and stability of the plasma column; as a further example the magnetic field out of plasma affects the operation of a number of critical components, included the diagnostic system and the neutral beam. Therefore the actual magnetic field has to be suitably calculated and his shape controlled within the required limits. Due to the complexity of its design, the problem is quite critical for the ITER project. In this paper the problem is discussed both from mathematical and numerical point of view. In particular, a complete formulation is proposed, taking into account both the presence of the non linear magnetic materials and the fully 3D geometry. Then the quality level requirements are discussed, included the accuracy of calculations and the spatial resolution. As a consequence, the numerical tools able to fulfil the quality needs while requiring reasonable computer burden are considered. In particular possible tools based on numerical FEM scheme are considered; in addition, in spite of the presence of non linear materials, the practical possibility to use Biot-Savart based approaches, as cross check tools, is also discussed. The paper also analyses the possible geometrical simplifications of the geometry able to make possible the actual calculation while guarantying the required accuracy. Finally the characteristics required for a correction system able to effectively counteract the magnetic field degradation are presented. Of course a number of examples will be also reported and commented. (author)