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The base plasma heat load on the ITER first wall results from the projection of the plasma scrape off layer power profile onto the shaped surface of the first wall panels. The actual heat load on the first wall is larger than the base heat load, due to enhancement factors such as design approximations, manufacturing and assembly errors, and deviations to the ideally toroidal magnetic field configurations. The heat load increase is accounted for through penalty coefficients, ranging individually from a few percent to 50%. The probabilistic combination of individual penalty coefficients to give a global penalty coefficient is presented in the paper. The global penalty coefficient ranges from 1.37 to 2.44, depending of the first wall row and plasma phase. [ABSTRACT FROM AUTHOR]

• Active wall monitoring systems are an increasingly important asset for safe operation of metallic wall steady state magnetic fusion experiments. • Wall Infrared diagnostic connection to plasma control system enables wall monitoring. • WEST operates the C4 experimental campaign in 2019 with an ITER relevant wall monitoring system (soft control plus hard interlock). • The control activates on 63 occurrences, principally on an upper divertor pipe being heated by particles losses. • It facilitated WEST path to high power operation during C4 campaign, by effectively managing the technical risks to critical wall components. A real time Wall Monitoring System (WMS) is used on the WEST tokamak during the C4 experimental campaign. The WMS uses the wall surface temperatures from 6 fields of view of the Infrared viewing system. It extracts the raw digital data from selected areas, converts it to temperatures using the calibration and write it on the shared memory network being used by the Plasma Control System (PCS). The PCS feeds back to actuators, namely the injected power from 5 antennae's of the lower hybrid and ion cyclotron resonance radiofrequency (RF) heating systems. WMS activates feed back control 63 times during C4, which is 14 % of the plasma discharges. It activates mainly as the result of a direct RF loss to the upper divertor pipes. The feedback control maintains the wall temperature within the operation envelope during 97 % of the occurrences, while enabling plasma discharge continuation. The false positive rate establishes at 0.2 %. WMS significantly facilitated the operation path to high power operation during C4, by managing the technical risks to critical wall components. [ABSTRACT FROM AUTHOR]

As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are available in ITER Organization on full scale prototypes of Shield Blocks #08 and #14, CFD analyses have been performed to investigate their fluid-dynamic behaviour under steady state conditions and compare the numerical and experimental results. Results obtained are presented and critically discussed. [ABSTRACT FROM AUTHOR]

Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ∼90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the 'W' Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (∼50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (∼35%) released in between discharges. Although a significant amount of N2(18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade. [ABSTRACT FROM AUTHOR]

This paper presents the method toward quantifying the global uncertainty on the blackbody temperature measurement from the infrared (IR) diagnostics of WEST. To address this goal, the main functional elements of the IR diagnostics are identified. Then the temperature calibration and calculation principles are presented and analysed to extract the main potential uncertainty contributors, such as the optical transmission coefficients and their stray lights, or the accuracy of temperature references used for calibration. These contributors are individually estimated, or experimentally measured when a supposed effect on the overall uncertainty is identified, like environmental conditions or parasitic radiations. In particular, effects of environmental temperature on the transmission lines and camera is thoroughly studied. All contributions are then aggregated in the uncertainty propagation calculation and results in an overall temperature uncertainty versus the temperature estimation. The uncertainty is in the range of 5–10% for blackbody temperatures above 200 °C, and progressively worsens when temperature decreases. [ABSTRACT FROM AUTHOR]

A tangential infrared thermography system is installed in the WEST tokamak to observe the thermal scene of the first wall and the divertor components through a wide-angle tangential view. The goal is to provide and monitor the heat load deposition on the plasma facing components. About one-sixth of the chamber is observed, including sectors of the baffle, the lower and upper divertors, one inner bumper, vertical displacement event (VDE) protections, one ICRH antenna, and other parts of the wall. It is used for real time machine protection by monitoring temperature thresholds in delimited region of interest, and for analysis of normal or specific heat load events during operation such as VDE, ELM, disruption or runaways. This wide-angle view uses one aspherical and one on-axis plane mirrors, a sapphire window and three lenses for the objective of the camera. The optical line is optimized for two wavelengths 1.7 and 4 µm. The field of view is 60° on a 512×640 pixels Focal Plane Array. The endoscope is fully actively cooled. The thermal scene is complex to interpret given the fully metallic and radiative environment and the uncertainties on the emissivity which is angular-dependent and changes with the surface properties. This involves significant inaccuracy on the recovery of the real temperature. In this context, an interpretation by modeling approach is better suited, based on ray-tracing simulations taking into account the optical properties of materials. For instance, this allowed discriminating reflections patterns from real thermal events in the wide-angle view. A description of the wide-angle infrared diagnostic and its performances is presented as well as experimental measurements obtained in the WEST Tokamak. [ABSTRACT FROM AUTHOR]

• Successful high heat flux test results of ITER First Wall mock-ups after neutron irradiation • High heat flux testing of actively cooled mock-ups manufactured with the relevant technologies in relevant geometry to the ITER First Wall, by the suppliers involved in the series production for ITER • High heat flux testing of neutron-irradiated beryllium flat tiles with hypervapotron heat sink • Details of high heat flux test by electron beam facility, JUDITH facility and details of neutron irradiation in the fission reactor, HFR Eight mock-ups imitating the ITER Enhanced Heat Flux First Wall panel were subjected to thermal fatigue test after neutron irradiation at two dose levels, ≈0.1 dpa and ≈0.5 dpa. All the mock-ups successfully withstood repeated thermal fatigue loads up to 4.7 MW/m2. The test results confirmed the performance of mock-ups in the tested condition, specifically the design and manufacturing technologies of the suppliers of two domestic agencies in charge of the series production. [ABSTRACT FROM AUTHOR]

Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed. [ABSTRACT FROM AUTHOR]

The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised. [ABSTRACT FROM AUTHOR]

This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m 2 is described in detail. [ABSTRACT FROM AUTHOR]