Numerical analysis of hydrogen risk mitigation measures for support of ITER licensing

In the ITER wet bypass scenario, water leakage, air ingress and hot dust (Be, W, and C) in the vacuum vessel could generate combustible hydrogen-air-steam mixture. Hydrogen combustion may threaten the integrity of the ITER VV and lead to radioactivity release. To prevent hydrogen energetic combustion, nitrogen injection system in VV and hydrogen recombination system in the pressure suppression tank (ST) were proposed. The main objectives of this analysis are to study the distribution of hydrogen-air-steam mixtures in the ITER sub-volumes, to investigate the feasibility of the nitrogen injection system to fully inert the atmosphere in the VV and to evaluate the capability and efficiency of the hydrogen recombination system to remove hydrogen in the ST. 3D computational fluid dynamics (CFD) code GASFLOW was used to calculate the evolution of the mixtures and to evaluate the hydrogen combustion risks in the ITER sub-volumes. The results indicate that the proposed hydrogen risk mitigation systems will generally prevent the risks of hydrogen detonation and fast deflagration. However, the atmosphere in ITER sub-volumes cannot be completely inerted at the early stage of the scenario. Slow deflagrations could still generate quasi-static pressures above 1 bar in the VV. The structural impact of the thermal and pressure loads generated by hydrogen combustions will be investigated in future studies.     

RADI—A RF source size-scaling experiment towards the ITER neutral beam negative ion source

IPP Garching is currently developing a negative hydrogen ion RF source for the ITER neutral beam system. The source demonstrated already current densities in excess of the ITER requirements (>200 A/m² D⁻) at the required source pressure and electron/ion ratio, but with only small extraction area and limited pulse length. A new test facility (RADI) went recently into operation for the demonstration of the required (plasma) homogeneity of a large RF source and the modular driver concept.The source with the dimension of 0.8 m × 0.76 m has roughly the width and half the height of the ITER source; its modular driver concept will allow an easy extrapolation in only one direction to the full size ITER source. The RF power supply consists of two 180 kW, 1 MHz RF generators capable of 30 s pulses. A dummy grid matches the conductance of the ITER source. Full size extraction is presently not possible due to the lack of an insulator, a large size extraction system and a beam dump.The main parameters determining the performance of this “half-size” source are the negative ion and electron density in front of the grid as well as the homogeneity of their profiles across the grid. Those will be measured by optical emission and cavity ring down spectroscopy, by Langmuir probes and laser detachment. These methods have been calibrated to the extracted current densities achieved at the smaller source test facilities at IPP for similar source parameters. However, in order to get some information about the possible ion and electron currents, local single aperture extraction with a Faraday cup system is planned.     

Numerical simulation of the combined effects of plasma heating and neutron heating loads on the ITER first wall

To understand the combined effect of plasma heating and neutron heating loadings, the distributions of temperature, stress, and strain in different two-dimensional first wall panel models under normal ITER operation condition were simulated using finite element method. The maximum temperature occurs at the Be armor, and reaches 461 °C. High thermal stresses (in the range of 80-200 MPa) are found at the interface between the Be armor and the CuCrZr layer. The maximum thermal stress reaches 324 MPa in the SS316L cooling tube (20 mm diameter), exceeding its yield strength and resulting in a maximum strain of about 1.7% at the tube inner surface. These simulation results are useful for the design and operation of ITER.     

Raman spectroscopy of carbon dust samples from NSTX

The Raman spectrum of dust particles exposed to the NSTX plasma is different from the spectrum of unexposed particles scraped from an unused graphite tile. For the unexposed particles, the high energy G-mode peak (Raman shift ∼ 1580 cm⁻¹) is much stronger than the defect-induced D-mode peak (Raman shift ∼ 1350 cm⁻¹), a pattern that is consistent with Raman spectrum for commercial graphite materials. For dust particles exposed to the plasma, the ratio of G-mode to D-mode peaks is lower and becomes even less than 1. The Raman measurements indicate that the production of carbon dust particles in NSTX involves modifications of the physical and chemical structure of the original graphite material. These modifications are shown to be similar to those measured for carbon deposits from atmospheric pressure helium arc discharge with an ablating anode electrode made from a graphite tile material. We also demonstrate experimentally that heating to 2000-2700 K alone cannot explain the observed structural modifications indicating that they must be due to higher temperatures needed for graphite vaporization, which is followed either by condensation or some plasma-induced processes leading to the formation of more disordered forms of carbon material than the original graphite.     

A preliminary assessment of beryllium dust oxidation during a wet bypass accident in a fusion reactor

A beryllium dust oxidation model has been developed at the Idaho National Laboratory (INL) by the Fusion Safety Program (FSP) for the MELCOR safety computer code. The purpose of this model is to investigate hydrogen production from beryllium dust layers on hot surfaces inside a fusion reactor vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). This beryllium dust oxidation model accounts for the diffusion of steam into a beryllium dust layer, the oxidation of the dust particles inside this layer based on the beryllium–steam oxidation equations developed at the INL, and the effective thermal conductivity of this beryllium dust layer. This paper details this oxidation model and presents the results of the application of this model to a wet bypass accident scenario in the ITER device.     

Interfacial properties of HIP joints between beryllium and reduced activation ferritic/martensitic steel

ITER test blanket modules are the most important components to validate energy production and fuel breeding for future fusion demonstration reactors. Reduced activation ferritic/martensitic steel is recognized as one of the promising structural materials for the breeding blanket systems. Beryllium is a primary candidate plasma facing materials for ITER blanket. In this work, the interfacial properties of Be/reduced activation ferritic/martensitic steel (RAF/Ms) joints were investigated for the first wall of an ITER test blanket module (TBM). The joints were produced by the solid-state hot isostatic pressing (HIP) method. Chromium (Cr) was used as a diffusion barrier with a thickness of 1 μm or 10 μm, formed by plasma vapor deposition on the Be surface. The HIPping was conducted at 1023 K and 1233 K with 160 MPa of static pressure. The temperatures are standard normalizing and tempering temperatures of F82H. EPMA showed the Cr layer effectively worked as a diffusion barrier at 1023 K. However, for the F82H/Be interface which underwent HIP at 1233 K followed by tempering a Be rich layer was formed. Bend tests revealed that a thin Cr layer and low temperature HIP is preferable. The joint with a thick Cr layer suffer from brittleness of Cr itself.     

Characterization of advanced cyanate ester/epoxy insulation systems before and after reactor irradiation

Insulation systems for fusion magnets have to operate in a harsh environment, especially also under intense radiation. Over the past years, cyanate ester (CE) resins have been playing an increasingly important role because of their enhanced temperature and radiation resistance compared to conventional epoxy resins. Blending CE with epoxy resins offers the possibility to manufacture radiation resistant insulations at a low price compared to pure CE materials. Therefore, it is of special interest to study the influence of the CE content and of the epoxy resin on the mechanical properties to find materials, which are suitable and economically reasonable for the specific demands of such magnets.In this study R-glass fiber/Kapton reinforced cyanate ester/epoxy blends with different CE content were investigated. Each material was exposed to conditions matching those expected for the ITER TF coil insulation as closely as possible. In order to characterize the mechanical properties, short-beam shear and static tensile tests were carried out at 77 K prior to and after irradiation to fast neutron fluences of up to 5 × 10²² m⁻² (E > 0.1 MeV), in the TRIGA reactor (Vienna) at ambient temperature (340 K). In addition, tension-tension fatigue measurements were performed in the load-controlled mode to simulate the pulsed operation conditions of ITER.     

Effect of radial hydrides on the axial and hoop mechanical properties of Zircaloy-4 cladding

The effect of radial hydrides on the mechanical properties of stress-relief annealed Zircaloy-4 cladding was studied. Specimens were firstly hydrided to different target hydrogen levels between 100 and 600 wt ppm and then thermally cycled in an autoclave under a constant hoop stress to form radial hydrides by a hydride reorientation process. The effect of radial hydrides on the axial properties of the cladding was insignificant. On the other hand, the cladding ductility measurements decreased as its radial hydride content increased when the specimen was tested in plane strain tension. A reference hydrogen concentration for radial hydrides in the cladding was defined for assessing the fuel cladding integrity based on a criterion of the tensile strength 600 MPa. The reference hydrogen concentration increased with the specimen (bulk) hydrogen concentration to a maximum of ∼90 wt ppm at the bulk concentration ∼300 wt ppm H and then decreased towards higher concentrations.     

PIXE and PIGE techniques for the analysis of Antarctic ice dust and continental sediments

An analytical procedure has been implemented in this work for an accurate geochemical characterization and quantitative analysis of the fine dust (particles diameter < 5 μm) trapped in Antarctic ice cores and the fine fraction of potential source areas (PSA) sediments by size selection, filtering and PIXE-PIGE combined measurements. The underestimation of concentrations of the lighter elements, like Na, Mg, Al and Si, due to X-ray self-absorption inside each individual aerosol particle, was also evaluated and the analytical overall accuracy tested by means of measurements performed on size selected certified mineral standards.     

Advances in optical thermometry for the ITER divertor

Thermography will be an important diagnostic on the ITER tokamak, but the inclusion of reflective materials such as tungsten in the design for ITER's first wall and divertor region presents problems for optical temperature measurement. The ongoing testing of ITER plasma facing components (PFCs) provides an excellent opportunity to resolve such problems. This has focused on the variation of PFC emissivity with temperature and time, as well as environmental influence on thermography. The sensitivity of these systems to ambient temperature, due primarily to modification of the transmission of the optical path, has been established and minimised. The accuracy of the system is then sufficient to measure the variation of emissivity in heated material samples, by comparing its front-face luminance measured with an infrared camera to the temperature given by an implanted thermocouple. Measurements on both tungsten and carbon fibre composite are in broad agreement with theory, and thus give the material's function of emissivity with temperature at the start of its life. To determine its evolution, a bicolour pyroreflectometer was then installed. This uses two lasers to measure the reflectivity in addition to the luminance at two wavelengths, and thus the true temperature can be calculated. This was validated against the instrumented sample, then used along with the camera to observe an ITER mock-up during ∼50,000 s of 5 MW/m² testing. Emissivity was seen to vary little in the 500 °C region. Higher temperature tests are ongoing.     

Investigations of single crystal and polycrystalline metal mirrors under erosion conditions in TEXTOR

Metal mirrors are planned for optical diagnostic systems of ITER. However, erosion, deposition and particle implantation can change the performance of mirrors. Mirrors made from the single crystal (SC) materials are among the main candidates for use in ITER diagnostic systems operating under erosion-dominated conditions. Laboratory tests have confirmed good optical performance of SC mirrors under erosion, but the dedicated direct comparative test in tokamak environment was missing.Such a direct test was performed in TEXTOR. Single crystal molybdenum, tungsten and polycrystalline (PC) molybdenum mirrors were exposed under the same conditions in the SOL plasma of TEXTOR. Surface and optical properties of mirrors were characterized before and after exposure. Before exposure glow discharge cleaning in hydrogen restored the reflectivity of mirrors oxidized during storage on air.No significant changes in total reflectivity were observed for all mirrors after exposure. Drastic increase of diffuse reflectivity was measured for PC Mo mirror, no change for a SC one. Thus, specular reflectivity of single crystal is higher than of polycrystalline one. The most affected wavelength range is 250-1000 nm, no significant change of reflectivity was noticed in the range 1000-2000 nm. Negligible effect of the exposure on polarization characteristics was observed.     

Tungsten behavior under proton flux and high temperature

An experimental study of the physico-chemical behavior of tungsten under severe conditions is presented. High temperatures (1300 ? T ? 2500 K) generated by concentrated solar energy, high vacuum (∼10⁻⁶ hPa) and proton flux (1 keV, ∼10¹⁷ ions m⁻² s⁻¹) have been applied on polycrystalline W samples to simulate expected and also unexpected high heat loads that can occur on the ITER divertor (nominal and accidental conditions). During experiment, in situ measurements are performed and the material degradation, the mass loss kinetics, the characterization of the different species coming from the materials under coupled proton flux and high temperatures and the optical properties (reflectivity) are followed. Material characterization using SEM and XRD was investigated before and after treatment to understand the observed behavior. Bidirectional reflectivity measurements were carried out on the tested samples to explain the surface modifications, between the reference sample, the heated sample and the heated and ion irradiated one that can act on the thermo-radiative properties of tungsten.     

Surface morphology influence on deuterium retention in beryllium films prepared by thermionic vacuum arc method

In a plasma-confinement device, material eroded from plasma facing components will be transported and re-deposited at other locations inside the reaction chamber. Since beryllium from the first wall of the ITER fusion reactor will be eroded, ionized in the scrape-off layer plasma and finally re-deposited on divertor surfaces flowing along the magnetic field, it is important to study the properties of divertor armour materials (C, W) coated with beryllium.By applying different bias voltages (−200 V to +700 V) to the substrates during deposition, the morphology of the obtained films was modified. The films’ morphology was characterized by means of AFM and SEM, and it was found that the coatings prepared using negative bias voltage at the substrate during deposition are more compact and have a smoother surface compared to the samples prepared with positive bias voltage. The thickness and composition of each film were measured using Rutherford backscattering spectrometry (RBS). A study of deuterium implantation and retention into the prepared films was performed at IPP Garching in the high current ion source.     

A remotely steered millimetre wave launcher for electron cyclotron heating and current drive on ITER

High-power millimetre wave beams employed on ITER for heating and current drive at the 170 GHz electron cyclotron resonance frequency require agile steering and tight focusing of the beams to suppress neoclassical tearing modes. This paper presents experimental validation of the remote steering (RS) concept of the ITER upper port millimetre wave beam launcher. Remote steering at the entrance of the upper port launcher rather than at the plasma side offers advantages in reliability and maintenance of the mechanically vulnerable steering system. A one-to-one scale mock-up consisting of a transmission line, mitre bends, remote steering unit, vacuum window, square corrugated waveguide and front mirror simulates the ITER launcher design configuration. Validation is based on low-power heterodyne measurements of the complex amplitude and phase distribution of the steered Gaussian beam. High-power (400 kW) short pulse (10 ms) operation under vacuum, diagnosed by calorimetry and thermography of the near- and far-field beam patterns, confirms high-power operation, but shows increased power loss attributed to deteriorating input beam quality compared with low-power operation. Polarization measurements show little variation with steering, which is important for effective current drive requiring elliptical polarization for O-mode excitation. Results show that a RS range of up to −12° to +12° can be achieved with acceptable beam quality. These measurements confirm the back-up design of the ITER ECRH&CD launcher with future application for DEMO.     

Development of water leak detection method in fusion reactors using water-soluble gas

A water leak detection method using krypton (Kr) as a water-soluble tracer has been proposed for fusion reactors with fully circulating the in-vessel cooling water. This method was targeted for applying to the International Thermonuclear Experimental Reactor (ITER), and the 10⁻³ Pa m³/s order of water leak valves were fabricated and connected to the water loop circuit. The water leaks were effused into the vacuum vessel evacuated by a cryopump and the water dissolved Kr was detected by a quadrupole mass spectrometer (QMS). When two leak valves with 1 m distance were attached to the test pipe with 30 °C heating, two distinct, the mass to charge number ratio (m/e) of 84 peak current rises caused by the water leak were successfully detected with the time interval of 39 s. On the other hand, the water accession length as a function of the traveling time was calculated by considering a natural convection flow caused by the 30 °C heating, where the traveling time was 44.6 s for the 1 m length. This means that the observed positional accuracy is 12.6%, based on the calculation. To enhance the positional accuracy, the detailed flow simulation is indispensable. This method can be applied to the ITER condition.     

Preliminary assessment for dust contamination of ITER In-Vessel Transporter

After plasma operations of ITER, radioactive dust will have accumulated in the vacuum vessel (VV). The In-Vessel Transporter (IVT) will be introduced into the VV to remove the shield blanket modules for maintenance or replacement and later reinstall them. The IVT itself also needs to undergo regular maintenance in the Hot Cell Facility (HCF). It is assumed that maintenance workers will be exposed to radioactive dust that has adhered to the surfaces of the IVT. In this study, the areas of the IVT that may be contaminated by dust are evaluated to assess the level of exposure to workers during maintenance work in the HCF. Decontamination processes for the IVT, such as a combination of vacuuming and brushing, were investigated and the dose rate after these processes was evaluated. Even though dust was removed from surfaces where decontamination was possible, the dose rate was very high at some assessment points. To decrease the dose rate in accordance with ALARA policy, a decontamination plan and a maintenance plan, which includes the removal of dust, a radiation shield system, and a reduction in working time are proposed.     

Saturation in deuterium retention of CFC graphite targets exposed to beryllium-seeded plasmas on PISCES-B

ITER strike-plates are foreseen to be of carbon-fiber-composite (CFC). In this study the CFC bulk deuterium retention in ITER-relevant conditions is investigated. DMS 701 (Dunlop) CFC targets were exposed to plasma in PISCES-B divertor plasma simulator. Samples were exposed to both pure deuterium plasma and beryllium-seeded plasma at high fluences (up to ) and high surface temperature (1070 K). The deuterium contents of the exposed samples have been measured using both thermal-desorption-spectrometry (TDS) during baking at 1400 K and ion beam nuclear reaction analysis (NRA). The total deuterium inventory has been obtained from TDS while NRA measured the deuterium depth distribution. In the analysed fluence range at target temperature of 1070 K, no fluence dependence was observed. The measured released deuterium is . In the case of target exposure with beryllium-seeded plasma no change in the released amount of deuterium was found. The deuterium concentration inside the samples is almost constant until the probed depth of ?m, except in the first 1 μm surface layer, where it is 5 times higher than in the bulk. No C erosion/redeposition was observed in the Be-seeded plasma cases. The measured retention, applied to 50 m² of ITER CFC surface, would imply a tritium saturated value of 0.3 gT, much lower than the ITER safety limit of 350 g.     

Conceptual design of the FAST load assembly

Fusion advanced studies torus (FAST) is a proposal for a satellite facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST is a compact (R₀ = 1.82 m, a = 0.64 m, triangularity δ = 0.4) machine able to investigate non-linear dynamics effects of alpha particle behaviours in burning plasmas [1], [2] and [5]. The project is based on a dominant 30 MW of ion cyclotron resonance heating (ICRH), 6 MW of lower hybrid (LH) and 4 MW of electron cyclotron resonance heating (ECRH). FAST operates at a wide range [3] and [4] of parameters, e.g., in high performance H-mode (B_T up to 8.5 T; I_P up to 8 MA) as well as in advanced Tokamak operation (I_P = 3 MA), and full non-inductive current scenario (I_P = 2 MA). Helium gas at 30 K is used for cooling the resistive copper magnets [6]. That allows for a pulse duration up to 170 s. To limit the TF magnet ripple ferromagnetic insert have been introduced inside the vacuum vessel (VV). Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI). Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plates material, and argon or neon as the injected impurities to mitigate the thermal loads.     

Characterization of hydrogen concentration in Zircaloy-4 using ultrasonic techniques

The relationship between hydrogen concentration precipitated as hydride particles and ultrasonic parameters, such as velocity and attenuation, was examined in Zircaloy-4 samples for potential applications in the Non-Destructive Test Field. Different amounts of hydrogen (up to 517 ppm) were introduced in the samples by gaseous charging. Ultrasonic attenuation measurements were performed with compressive waves at frequencies of 10 and 30 MHz, and propagation velocity measurements were performed at 10 MHz. Ultrasonic velocity showed an approximately linear increase with hydrogen concentration and it could be used as an assessment parameter when the hydrogen level is high enough. Attenuation versus hydrogen concentration has been fitted by a logarithmic equation at 10 MHz. At 30 MHz a fluctuating behavior of the attenuation prevented measurement of the hydrogen concentration.