| 1. |
- Balbinot, L., et al.
(author)
-
Multi-code estimation of DTT edge transport parameters
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 34
-
Journal article (peer-reviewed)abstract
- The main goal of the Divertor Tokamak Test facility (DTT) is to operate with a high value of power-exhaust-relevant parameter Psoz/R in plasma scenarios similar to those foreseen for the Demonstration Fusion Power Plant (DEMO) in terms of low collisionality and neutral opacity. For these unique characteristics, accurate modelling of the principal scenario is necessary for machine designing. In edge numerical codes, cross-field transport profiles have a high impact on modelling results. This work aims at providing a coherent set of transport parameters for DTT full-power (FP) single-null (SN) scenario edge modelling. To evaluate such parameters for DTT, a transport analysis on the current machine has been performed using SOLEDGE2D-EIRENE and SOLPS-ITER. The transport parameters to be used in the simulations of the DTT single-null scenario were selected using two complementary methods. The first is the modelling of JET and Alcator C-Mod (C-Mod) with SOLEDGE2D-EIRENE and SOLPS-ITER, validating transport parameters by comparing modelling results to experimental data from pulses which are considered DTT-relevant. JET pulses were selected with the highest auxiliary input power (from 29 to 33 MW), plasma current and toroidal field to better match DTT parameters; nitrogen and neon seeded pulses were selected to check possible seeding material dependencies. The considered C-Mod pulse better matches DTT plasma density and neutral opacity. Transport parameters are then scaled to DTT according to scaling laws. The second method derives the transport parameters by tuning their values inside the DTT separatrix to reproduce the pedestal profiles predicted by the EPED model via the Europed code and applied in DTT. The derived set of DTT transport parameters is consistent with the results obtained by modelling present machines, reproduces the expected heat flux decay length in detached conditions and, inside the separatrix, reproduces the predicted pedestal using transport parameters which are coherent with what is predicted by the quasi-linear turbulent model QuaLiKiz.
|
|
| 2. |
- Corre, Y., et al.
(author)
-
Testing of ITER-grade plasma facing units in the WEST tokamak: Progress in understanding heat loading and damage mechanisms
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Journal article (peer-reviewed)abstract
- Assessing the performance of the ITER design for the tungsten (W) divertor Plasma Facing Units (PFUs) in a tokamak environment is a high priority issue to ensure efficient plasma operation. This paper reviews the most recent results derived from experiments and post-mortem analysis of the ITER-grade PFUs exposed in the WEST tokamak and the associated modelling, with a focus on understanding heat loading and damage evolution. Several shaping options, sharp or chamfered leading edge (LE), unshaped or shaped blocks with a toroidal bevel as foreseen in ITER, were investigated, under steady state heat fluxes of up to 120 MW⋅m−2 and 6 MW⋅m−2 on the sharp LE and top surface of the block, respectively. A very high spatial resolution (VHR) infrared (IR) camera (0.1 mm/pixel) was used to derive the temporal and surface distribution of the temperature and heat load on the castellated tungsten blocks for different geometric alignment and plasma conditions. Photonic modelling was required to reproduce the IR measurements in particular in the toroidal and poloidal gaps of the mono-block (MB) stacks where high apparent temperatures are observed. Specular reflection is found to be the dominant emitter in these parts of the blocks. W-cracking was observed on the leading edge of the blocks already within the first phase of plasma operation, during which the divertor was equipped with unshaped PFUs, including some intentionally misaligned blocks. Numerical analysis taking into account softening processes and mechanical stresses, revealed brittle failure due to transients as the dominant failure mechanisms. Ductile failure was observed in one particular block used for the melting experiment, therefore under extremely high steady state heat load conditions. W-melting achieved on actively cooled PFU exhibits specific features: shallow melting and slow melt displacement. Plasma exposure of pre-damaged PFUs at various damage levels (crack network and melted droplets) was carried out under high heat load conditions with several hours of cumulated plasma duration. IR data and preliminary surface analyses show no evidence of significant degradation damage progression under these conditions.
|
|
| 3. |
- Cupak, C., et al.
(author)
-
Absence of synergistic effects in quasi-simultaneous sputtering of tungsten by Ar and D ions
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 35
-
Journal article (peer-reviewed)abstract
- A quartz crystal microbalance was used to experimentally study the erosion of tungsten during rapidly alternating bombardment with 2 keV argon and deuterium projectiles. A key goal was to investigate whether the mean sputtering yield of the alternating irradiation can be predicted from data for sputtering yields of single ion species. In addition, influences by residual gas pressure in the UHV experiment and variable ion fluxes have been studied. Our results show that the mean sputtering yield of irradiations with alternating ion species can be well predicted for a range of different fluence ratios as a simple superposition of individual sputtering yields, weighted by the respective relative fluences. This finding supports that no synergistic sputtering effects were relevant in the investigated low-flux regime.
|
|
| 4. |
- De Angeli, M., et al.
(author)
-
Post-mortem and in-situ investigations of magnetic dust in ASDEX Upgrade
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 36
-
Journal article (peer-reviewed)abstract
- Pre-plasma mobilization of magnetic dust can be an important issue for future fusion reactors where plasma breakdown is critical. A combined on-line and off-line study of magnetic dust in ASDEX Upgrade is reported. Post-mortem collection revealed similar composition and morphology compared to other tokamaks, but the overall amount was much smaller. Optical and IR camera diagnostics excluded dust flybys prior to plasma start-up. The negative detection is discussed in light of the magnetic dust properties, the strength of mobilizing forces and the temporal evolution of the magnetic field.
|
|
| 5. |
- Dittrich, Laura, et al.
(author)
-
Impact of ion irradiation and film deposition on optical and fuel retention properties of Mo polycrystalline and single crystal mirrors
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Journal article (peer-reviewed)abstract
- Polycrystalline (PC) and single crystal (SC) molybdenum mirrors were irradiated with 98Mo+, 1H+, 4He+, 11B+ and 184W+. Energies were chosen to impact the optically active region (up to 30 nm deep) of Mo mirrors. Some surfaces were coated by magnetron sputtering either with B or W films 4–65 nm thick. The overall objective was to simulate the neutron-induced damage and transmutation (H, He), and the impact of H, He, B, W on the optical performance of test mirrors, and on fuel retention. In parallel, a set of PC Mo mirrors irradiated with 1.6 MeV 98Mo3+ to a damage of 2 dpa and 20 dpa was installed in the JET tokamak for exposure during deuterium-tritium campaigns. Data from spectrophotometric, ion beam and microscopy techniques reveal: (i) the irradiation decreased specular reflectivity, whereby the differences between PC and SC in reflectivity are very small, (ii) He is retained in bubbles within 25–30 nm of the subsurface layer in all irradiated materials, (iii) W, either deposited or implanted, decreases reflectivity, but the strongest reflectivity degradation is caused by B deposition. Laboratory studies show the correlation of damage and H retention. Several cycles of W deposition and its removal from SC-Mo mirrors by plasma-assisted methods were also performed.
|
|
| 6. |
- Fellinger, Joris, et al.
(author)
-
Tungsten based divertor development for Wendelstein 7-X
- 2023
-
In: Nuclear Materials and Energy. - 2352-1791. ; 37
-
Journal article (peer-reviewed)abstract
- Wendelstein 7-X, the world’s largest superconducting stellarator in Greifswald (Germany), started plasma experiments with a water-cooled plasma-facing wall in 2022, allowing for long pulse operation. In parallel, a project was launched in 2021 to develop a W based divertor, replacing the current CFC divertor, to demonstrate plasma performance of a stellarator with a reactor relevant plasma facing materials with low tritium retention. The project consists of two tasks: Based on experience from the previous experimental campaigns and improved physics modelling, the geometry of the plasma-facing surface of the divertor and baffles is optimized to prevent overloads and to improve exhaust. In parallel, the manufacturing technology for a W based target module is qualified. This paper gives a status update of project. It focusses on the conceptual design of a W based target module, the manufacturing technology and its qualification, which is conducted in the framework of the EUROfusion funded WPDIV program. A flat tile design in which a target module is made of a single target element is pursued. The technology must allow for moderate curvatures of the plasma-facing surface to follow the magnetic field lines. The target element is designed for steady state heat loads of 10 MW/m2 (as for the CFC divertor). Target modules of a similar size and weight as for the CFC divertor are assumed (approx. < 0.25 m2 and < 60 kg) using the existing water cooling infrastructure providing 5 l/s and roughly maximum 15 bar pressure drop per module. The main technology under qualification is based on a CuCrZr heat sink made either by additive manufacturing using laser powder bed fusion (LPBF) or by uniaxial diffusion welding of pre-machined forged CuCrZr plates. After heat treatment, the plasma-facing side of the heat sink is covered by W or if feasible by the more ductile WNiFe, preferably by coating or alternatively by hot isostatic pressing W based tiles with a soft OFE-Cu interlayer. Last step is a final machining of the plasma-exposed surface and the interfaces to the water supply lines and supports to correct manufacturing deformations.
|
|
| 7. |
- Kit, A., et al.
(author)
-
Developing deep learning algorithms for inferring upstream separatrix density at JET
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 34
-
Journal article (peer-reviewed)abstract
- Predictive and real-time inference capability for the upstream separatrix electron density, ne, sep, is essential for design and control of core-edge integrated plasma scenarios. In this study, both supervised and semi -supervised machine learning algorithms are explored to establish direct mapping as well as indirect compressed representation of the pedestal profiles for predictions and inference of ne, sep. Based on the EUROfusion pedestal database for JET (Frassinetti et al., 2021), a tabular dataset was created, consisting of machine parameters, fraction of ELM cycle, high resolution Thomson scattering profiles of electron density and temperature, and ne, sep for 608 JET shots. Using the tabular dataset, the direct mapping approach provides a mapping of machine parameters and ELM percentage to ne, sep. Through representation learning, a compressed representation of the experimental pedestal electron density and temperature profiles is established. By conditioning the representation with machine control parameters, a probabilistic generative predictive model is established. For prediction, the machine parameters can be used to establish a conditional distribution of the compressed pedestal profiles, and the decoder that is trained as part of the algorithm can be used to decode the compressed representation back to full pedestal profiles. Although, in this work, a proof-of-principle for predicting and inferring ne, sep is given, such a representation learning can be used also for many other applications as the full pedestal profile is predicted. An implementation of this work can be found at https://github.com/ fusionby2030/psi_2022.
|
|
| 8. |
- Lindgren, Kristina, 1989, et al.
(author)
-
Elemental distribution in a decommissioned high Ni and Mn reactor pressure vessel weld metal from a boiling water reactor
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; , s. 101466-101466
-
Journal article (peer-reviewed)abstract
- In this paper, weld metal from unique material of a decommissioned boiling water reactor pressure vessel is investigated. The reactor was in operation for 23 effective full power years. The elemental distribution of Ni, Mn, Si and Cu in the material is analysed using atom probe tomography. There are no well-defined clusters of these elements in the weld metal. However, some clustering tendencies of Ni was found, and these are interpreted as a high number density of small features. Cu atoms were found to statistically be closer to Ni atoms than in a fully random solid solution. The impact of the non-random elemental distribution on mechanical properties is judged to be limited.
|
|
| 9. |
- Mishchenko, Yulia, et al.
(author)
-
Thermophysical properties and oxidation behaviour of the U0.8Zr0.2N solid solution
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier Ltd. - 2352-1791. ; 35
-
Journal article (peer-reviewed)abstract
- Thermophysical properties and oxidation behaviour of the composite pellet UN–20 vol%ZrN were investigated experimentally and compared with the behaviour of the pure UN pellet. A compound of a single phase, a solid solution of the average composition U0.8Zr0.2N, was obtained by Spark Plasma Sintering (SPS) of the powders UN and ZrN. Crystallographic and microstructural characterisation of the composite was performed using Scanning Electron Microscopy (SEM), standardised Energy Dispersive Spectroscopy (EDS) and Electron Backscatter Diffraction (EBSD). Nano hardness and Young's modulus were also measured by the nanoindentation method. High-Temperature X-ray diffraction (XRD) was applied to obtain the lattice expansion as a function of temperature (room temperature to 673 K). Thermogravimetric Analysis (TGA) was applied to evaluate oxidation behaviour in air. Results demonstrate that the fabrication method results in a matrix of solid solution with homogeneous composition averaged to U0.8Zr0.2N. The mechanical properties of such solution are uniform, with variation only due to the crystallographic orientation of the grains of the solution phase, similar to pure UN. The obtained value for the average linear thermal expansion coefficient is α¯ = 7.94 × 10-6/K, which compares well to UN (α¯ = 7.95 × 10-6/K) for the same temperature range. The degradation behaviour of the composite pellet UN-20 vol%ZrN in air shows a lower oxidation onset temperature, compared to pure UN, with the final product of oxidation being mainly U3O8. Smaller crystallites in the product of corrosion of the composite pellet indicate that the mechanism of degradation of the solid solution phase U0.8Zr0.2N is accompanied by the formation of two distinct oxides and their interaction.
|
|
| 10. |
- Paschalidis, Konstantinos, et al.
(author)
-
Melt dynamics with MEMENTO — Code development and numerical benchmarks
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Journal article (peer-reviewed)abstract
- The new numerical implementation of the MEMOS-U physics model, fully validated in multiple EUROfusion sponsored experiments, is presented. The computational tool - MEMENTO (MEtallic Melt Evolution in Next-step TOkamaks)- is able to address fusion-relevant melting scenarios that feature complex plasma-facing component geometries, involve intricate plasma wetting patterns and are characterized by vast spatio-temporal scale separations. The high level architecture of the code is discussed and numerical benchmarks of the heat transfer, fluid dynamics and current propagation solvers are presented.
|
|
| 11. |
- Patnaik, Sobhan, et al.
(author)
-
Crystallographic characterization of U 2 CrN 3 : A neutron diffraction and transmission electron microscopy approach
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 35
-
Journal article (peer-reviewed)abstract
- In this study, neutron diffraction and transmission electron microscopy (TEM) have been implemented to study the crystallographic structure of the ternary phase U2CrN3 from pellet to nano scale respectively. Recently microstructural evaluation of this ternary phase has been performed for the first time in pellet condition, overcoming the Cr evaporation issue during the conventional sintering process. In this work for the first time, the crystallographic structure of the ordered ternary U2CrN3 phase, stabilized in pellet condition, has been obtained by implementing neutron diffraction. For this study, pellets of the composite material UN with 20 vol% CrN were fabricated by powder metallurgy by mixing UN and CrN powders followed by Spark Plasma Sintering (SPS). TEM was used to investigate the nanoscale structure with a thin lamella of the order of 100–140 nm produced by focused ion beam (FIB). The neutron data revealed the phase composition of the pellet to be primarily 54(8) wt.% U2CrN3, in good agreement with the stoichiometry of starting reagents (UN and CrN powder) and metallographic analysis. Neutron data analysis confirms that all the crystallographic sites in U2CrN3 phase are fully occupied reinforcing the fully stoichiometric composition of this phase, however, the position of the N at the 4i site was found to be closer to the Cr than previously thought. TEM and selected area electron diffraction rendered nano-level information and revealed the presence of nano domains along grain boundaries of UN and U2CrN3, indicating a formation mechanism of the ternary phase, where the phase likely nucleates as nano domains in UN grains from migration of Cr.
|
|
| 12. |
- Petersson, Christopher, et al.
(author)
-
Slow strain rate testing of Fe-10Cr-4Al ferritic steel in liquid lead and lead-bismuth eutectic
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 34
-
Journal article (peer-reviewed)abstract
- The susceptibility of Fe-10Cr-4Al steel to liquid metal embrittlement (LME) in low oxygen liquid lead and lead-bismuth eutectic (LBE) environments has been investigated using a newly developed slow strain rate testing (SSRT) technique that can be employed at elevated temperatures. This study showed that the Fe-10Cr-4Al steel suffered embrittlement when exposed to LBE over a wide temperature range. The embrittlement, here measured as a reduction in fracture strain, was observed at the melting temperature of LBE and reached a maximum at approximately 375 degrees C. At temperatures above 425 degrees C, the material's ductility regained its original levels. The exposures in liquid lead showed no indications of embrittlement, but a ductile behavior over the entire tem-perature range studied (340-480 degrees C).
|
|
| 13. |
- Pitthan, Eduardo, et al.
(author)
-
Influence of thermal annealing and of the substrate on sputter-deposited thin films from EUROFER97 on tungsten
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 35
-
Journal article (peer-reviewed)abstract
- The modification of sputter-deposited films from EUROFER97 on tungsten during and after annealing were investigated in-situ and ex-situ. The annealing resulted in a densification of the film, formation of large grains, segregation of W at the surface, and the formation of Fe-W compounds at the interfacial region. Similar structural modifications were observed also for a film annealed on a MgO substrate, with an exception to the change in composition (no increase of W concentration). Results indicate that the substrate significantly affects thermally induced modifications of re-deposited EUROFER97.
|
|
| 14. |
- Pitthan, Eduardo, et al.
(author)
-
Thin films sputter-deposited from EUROFER97 in argon and deuterium atmosphere : Material properties and deuterium retention
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier. - 2352-1791. ; 34
-
Journal article (peer-reviewed)abstract
- Sputter-deposited thin films (33–1160 nm) from EUROFER97 were obtained on different substrates (C, Si, W, MgO) in argon and a mix of argon and deuterium atmosphere. The composition, microstructure, and mechanical properties of the films were analyzed and compared to those of the bulk material. The films feature lower density (-10%), higher hardness (+79%), and smaller crystallites in comparison to the bulk. Despite such differences, the elemental atomic composition of the films and the bulk was very similar, as determined by ion beam analysis. Deposition in deuterium-containing atmosphere resulted in a low deuterium incorporation (0.28% of atomic content), indicating low retention of hydrogen-isotopes in the deposited material.
|
|
| 15. |
- Shams-Latifi, Jila, et al.
(author)
-
Experimental electronic stopping cross-section of tungsten bulk and sputter-deposited thin films for slow protons, deuterons and helium ions
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 36
-
Journal article (peer-reviewed)abstract
- The experimental electronic stopping cross-section of tungsten for low-energy protons, deuterons, and helium ions is deduced from backscattering experiments from thin films and bulk using time-of-flight low-energy ion scattering (ToF-LEIS). Two complementary experimental approaches showed consistent results in the energy ranges of 0.3-10 keV for protons, 0.33-10 keV for deuterons, and 0.7-10 keV for He+ ions. In relative mea-surements, a Au sample was used as the reference, while in absolute energy loss measurements, sputter-deposited thin films of tungsten on carbon substrates were employed. The experimental energy-converted spectra were compared to Monte-Carlo simulations in both approaches for quantitative analysis taking the influence of plural and multiple scattering into account. The results show proportionality to the ion velocity. We discuss the present datasets in comparison to semiempirical modelling and predictions from theory.
|
|
| 16. |
- Simpson, J., et al.
(author)
-
Investigation of the dependence of p(e,ped) on n(e,sep) in JET H-Mode plasmas using integrated JETTO-MISHKA-FRANTIC simulations
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 34
-
Journal article (peer-reviewed)abstract
- Experimentally, it has been observed in high-confinement (H-Mode) plasmas with Edge Localised Modes (ELMs) on JET that the pressure pedestal (p(e,ped)) is degraded by approximately a factor of two when there is a change in electron separatrix density, n(e,sep), from 1 - 4 x 10(19) m(-3). Previous work using the pedestal stability code EUROPED, has been able to predict the degradation of p(e,ped) but only for n(e,sep) = 1.5 x 10(19) m(-3). In this work, we apply a coupled code JETTO-MISHKA-FRANTIC, to self-consistently predict the transport in the pedestal region and neutral source with varying separatrix conditions. The code feeds back on the transport in the pedestal region to achieve profiles that are marginally stable to ideal MHD modes (continuous ELM model in JETTO). When accounting for the change in electron separatrix temperature (T-e,T-sep), ion separatrix temperature (T-e,T-sep) and the poloidally integrated neutral flux crossing the separatrix (Gamma(sep,neui)) as it changes with n(e,sep) (according to a scan in n(e,sep) in the edge code EDGE2D-EIRENE), no degradation in p(e,ped) was observed in JETTO-MISHKA-FRANTIC in contrast to experiment. Instead, an increase in p(e,ped) with n(e,sep) was observed which is driven by an increasing density pedestal (n(e,ped)). Within the presented JETTO-MISHKA-FRANTIC simulations, changing the pedestal width by a factor of two and a half in normalised poloidal flux (psi(n)) resulted in an approximately 40% degradation in p(e,ped) for n(e,sep) = 1 - 3 x 10(19) m(-3). This change in pedestal width was not supported by experimental data. A scan in the ratio of particle and energy transport in the pedestal (D/chi) was found to have a negligible effect on p(e,ped). Qualitative agreement between JETTO-MISHKA-FRANTIC with EUROPED was found when the input density profiles are identical.
|
|
| 17. |
- Sweidan, Faris, et al.
(author)
-
Finite element modeling of UN-UO2 and UN-X-UO2 (X=Mo, W) composite nuclear fuels : temperature-dependent thermal conductivity and fuel performance
- 2023
-
In: Nuclear Materials and Energy. - 2352-1791.
-
Journal article (peer-reviewed)abstract
- In this study, the temperature-dependent effective thermal conductivity of the innovative UN-X-UO2 (X=Mo, W) nuclear fuel composite has been estimated in the temperature range from room temperature to 2000 K. This composite fuel concept is considered as a promising accident tolerant fuel for light water reactors (LWRs). Following the previously reported experimental composite design, the composite fuel thermal conductivity was calculated using Finite Element modeling (FEM), and it is compared with analytical models of thermal conductivity for 10, 30, 50, and 70 wt.% uncoated/coated UN microspheres in a UO2 matrix. The FEM results show an expected increase in the fuel thermal conductivity as the wt.% of the coated/uncoated UN microspheres increases – from 1.5 to 5.7 times the UO2 reference at 2000 K. However, the analytical models show an overestimation of the fuel thermal conductivity as the wt.% increases. The results also show that Mo and W coatings have similar thermal behaviors and the coating thickness varying from 1-5 μm has an insignificant effect on the thermal behavior of the composite. However, at higher weight fractions, the thermal conductivity of the fuel composite at room temperature is substantially influenced by the high thermal conductivity coatings exceeding that of UN. Thereafter, the thermal conductivity profiles from FEM were used in the fuel thermal performance evaluation during LWR normal operation to calculate the maximum centerline temperature of the fuel composites. The results show a significant decrease in the fuel maximum centerline temperature ranging from −72 K for 10 wt.% UN to −438 K for 70 wt.% UN compared to the UO2 under the same irradiation conditions, providing an enhanced safety margin and thermal and neutronic advantages.
|
|
| 18. |
- Tichit, Q., et al.
(author)
-
Infrared detection of tungsten cracking on actively cooled ITER-like component during high power experiment in WEST
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Journal article (peer-reviewed)abstract
- The consequences of tungsten (W) damaging processes, such as cracking and melting, on divertor lifetime and plasma operation are high priority issues for ITER. A sustained melting experiment was conducted in WEST using a 2 mm deep groove geometry on the upstream mono-block (MB) to overexpose the sharp leading edge (LE) of the downstream MB. W-cracking has been evidenced for the first time with a very high spatial resolution infrared camera before tungsten melting was reached. These cracks develop when the monoblock temperature is about 2600 degrees C, thus higher than both ductile to brittle transition and softening threshold of tungsten, suggesting that these cracks are different from the ones observed in previous campaigns where brittle failure was involved, because of transient events on cold monoblock. Post-exposure analyses have been performed on the damaged monoblock, highlighting 12 main cracks on the LE, with a width varying from 33 mu m to 77 mu m, and an average spacing of 0.45 mm. Parallel heat flux about 90 MW/m2 has been derived from infrared temperature measurements, with a heat flux decay length on the target of 4 mm. The T-REX modelling code suggest here that with these thermal inputs, a crack can initiates due to thermal cycling without disruption, with a ductile failure, under 1 to 5 cycles for a tungsten DBTT varying from 400 degrees C to 500 degrees C.
|
|
| 19. |
- Vignitchouk, Ladislas, et al.
(author)
-
Instability of molten beryllium layers during ITER thermal quenches
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Journal article (peer-reviewed)abstract
- The production and dynamics of beryllium melt pools are simulated in conditions relevant to unmitigated thermal quenches in ITER. Rayleigh-Taylor instabilities fed by Lorentz forces due to induced eddy currents are found to result in significant material losses from droplet ejection, corresponding to equivalent eroded depths up to 500 mu m. Different thermal and electromagnetic loading scenarios are investigated, demonstrating a strong dependence of wall damage on the intensity of the heat and current pulses. The contribution of convection flows stemming from surface tension gradients along the plasma-liquid interface is also elucidated.
|
|
| 20. |
- Zayachuk, Y., et al.
(author)
-
Impact of water ingress on deuterium release, oxidation, and dust generation in beryllium plasma-facing components
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 35
-
Journal article (peer-reviewed)abstract
- Beryllium samples from the JET ITER-like wall limiter tiles with either co-deposits or surface cracks caused by melt damage, were immersed into boiling water for 4 h 15 min to simulate and assess the impact of coolant water ingress into a tokamak on the state of Be components. Microscopy of the water-treated surfaces and the lack of residue in the water revealed that no thermomechanical damage (cracking or exfoliation) occurred to the samples during the exposure. Ion beam analysis showed no measurable release of deuterium from the samples. Combined ion beam analysis and Raman spectroscopy indicated only some degree of surface oxidation, but no thick oxide films were formed.
|
|
