Tritium Retention Research Articles

An integral approach to plasma-wall interaction modelling for EU-DEMO

Abstract The integral approach to plasma-wall interaction (PWI) modelling for DEMO is presented, which is a part of EUROfusion Theory and Advanced Simulation Coordination (E-TASC) activities that were established to advance the understanding and predictive capabilities for modelling of existing and future fusion devices using a modern advanced computing approach. In view of DEMO design, the aim of PWI modelling activities is to assess safety-relevant information regarding erosion of plasma-facing components (PFC), including its impact on plasma contamination, dust production, fuel inventory, and material response to transient events. This is achieved using a set of powerful and validated computer codes that deal with particular PWI aspects and interact with each other by means of relevant data exchange. Steady state erosion of tungsten PFC and subsequent transport and re-deposition of eroded material are simulated with the ERO2.0 code using a DEMO plasma background produced by dedicated SOLPS-ITER simulations. Dust transport simulations in steady state plasma also rely on the respective SOLPS-ITER solution and are performed with the MIGRAINe code. On the way to improved simulations of tungsten erosion in the divertor of DEMO, relevant high density sheath models are being developed based on Particle-in-Cell (PIC) simulations with the state-of-the-art BIT code family. PIC codes of the SPICE code family, in turn, provide relevant information on multi-emissive sheath physics, such as semi-empirical scaling laws for field-assisted thermionic emission. These scalings laws are essential for simulations of material melting under transient heat loads that are performed with the recently developed MEMENTO code, the successor of MEMOS-U. Fuel retention simulations assess tritium retention in tungsten and structural materials and permeation to coolant accounting for neutron damage, in particular for divertor monoblocks of different sizes, using the FESTIM code, and for the first wall, in particular in view of tritium self-sufficiency, using the TESSIM code. Respective code-code dependencies and interactions, as well as modelling results achieved to date are discussed in the contribution.

Resolving studies of Balmer alpha lines relevant to the LIBS analysis of hydrogen isotope retention

Utilizing Laser-Induced Breakdown Spectroscopy (LIBS) for detecting deuterium and tritium retention in fusion devices poses a significant challenge due to the experimental limitations in resolving hydrogen isotope Balmer alpha lines (Hα, Dα, and Tα). This study utilizes the Rayleigh criterion to distinguish Tα and Dα lines by determining the maximum line widths and corresponding plasma parameters. Experimental validation was performed through LIBS analysis of heavy water-doped graphite/silica gel targets in both argon and helium atmospheres to assess the predicted plasma parameters and line profile shapes. The optimization of laser pulse energy, gas pressure, delay, and gate times aimed at achieving fully resolved lines based on the intensity, width, and the dip between deuterium and hydrogen Balmer alpha lines. By fine-tuning these experimental parameters, the study successfully achieved a dip of less than 10 % between the Hα and Dα lines with a satisfactory signal-to-noise ratio, demonstrating the feasibility of fully resolving the Tα and Dα lines. These findings underscore the potential of LIBS in enhancing the detection of deuterium and tritium retention in fusion devices.

Boron powder injection experiments in WEST with a fully actively cooled, ITER grade, tungsten divertor

Reactor relevant fusion devices will use tungsten (W) for their plasma facing components (PFCs) due to its thermomechanical properties and low tritium retention. However, W introduces high-Z impurities into the plasma, degrading its performance. Different wall conditioning methods have been developed to address this issue, including coating of W PFCs with layers of low-Z material. Wall conditioning by boron (B) powder injection using an impurity powder dropper (IPD) is being studied in WEST. Two series of experiments were conducted since the installation of the new ITER grade full W divertor. During the first series in 2023 ∼ 1 g of B powder was injected in total at a maximum rate of ∼ 58 mg/s, both of which are three times greater than respective values in the initial WEST powder injection experiments. The second series of experiments included injection of B and BN powders for comparison of their effects on plasma performance. The presence of an instantaneous conditioning effect is suggested by visible spectroscopy measurements of low-Z impurity lines and a rollover of total radiated power past an injection rate of ∼ 20 mg/s was observed. Presence of B coating layer formation is supported by the evolution of the average radiance of visible lines of B, W and oxygen (O). To understand B transport, an interpretative modeling workflow is employed, utilizing the SOLEDGE-EIRENE fluid boundary plasma code and the Dust Injection Simulator (DIS) code. Parameters like B perpendicular diffusivity and recycling coefficients are varied to match experimental results to see if the initial assumption of B sticking to the PFCs immediately after the contact with the wall is adequate for correctly modelling its distribution on the PFCs.

Studies of Hydrogen Desorption in Graphite Using Extreme Time Scale Chronoamperometry in a Molten Salt Environment

Controlling tritium is a major issue in developing nuclear reactors which employ molten salts; to aid this, we must understand tritium retention in these reactors. An important factor that must be considered for tritium retention is the effect the molten salt environment has on the interactions between tritium and the graphite moderator. Traditionally, Thermal Desorption Spectroscopy (TDS) is used to study the adsorption of hydrogen and its isotopes onto graphite; however, this is done in a vacuum at a constant heating rate to remove desorbed hydrogen where the gas is then measured for hydrogen gas concentration vs. temperature. To test in a molten salt environment, a method is needed to create a concentration gradient to drive diffusion of hydrogen to the surface of the graphite since a vacuum cannot be applied. This can be done electrochemically using a graphite working electrode submerged in salt by applying a constant oxidizing potential. Desorbed hydrogen gas is oxidized by this potential on the surface of the graphite, creating a concentration gradient which continuously drives diffusion as the graphite and salt is heated. This technique, known as Electrochemical Thermal Desorption Spectroscopy (ETDS) can be used to study tritium interactions with graphite in molten salt using hydrogen as a surrogate.Previous attempts of ETDS used chronoamperometry with a constant heating rate to determine current peaks and relate them to hydrogen desorption peaks from various traps in graphite identified in TDS literature. The largest problem facing this previous work was the contribution to measured current from sources other than hydrogen oxidation at the surface. Work has been done to use a “blank” sample of graphite to run heated chronoamperometry and act as a baseline to subtract from the “charged” sample data to find current contributions only from hydrogen oxidation. Additionally, characterization of the salt is being done to determine a theoretical background current based on the kinetics of possible reactions and the effects of overpotential, concentration and high temperature.Current challenges facing ETDS development are being studied. In this work, we examine the contributions of background reactions to measured current, as well as current contributions due to the oxidation of hydrogen within the electrode rather than at the surface. The answers to these questions will increase signal to noise ratios of ETDS, allowing the technique to extract valuable kinetic parameters of hydrogen interactions and quantify hydrogen in graphite in a molten salt environment.

Introduction on tritium transport analysis model for HCCP breeding blanket system

In the context of implementing the He-cooled breeding blanket system for a fusion reactor, several crucial aspects need consideration. Among these, the development of a tritium transport model plays a pivotal role in ensuring safety and effective design. Given that tritium release into the environment from the breeding blanket system poses a radioactive risk, accurate calculation and prediction area essential to prevent potential incidents in a fusion reactor. To address this challenge, a collaborative effort between the Korea Institute of Fusion Energy (KFE) and the University of California, Los Angeles (UCLA) resulted in the creation of THETA-FR (Tritium/Hydrogen Enhanced dynamic Transport Analysis Tool for Fusion Reactor). THETA-FR is an integrated analysis tool that combines Matlab Simulink and COMSOL Multiphysics. Its purpose is to model and predict the dynamic transport phenomena of H isotopes (H/D/T) within the breeding blanket system. Specifically, THETA-FR focuses on transient tritium retention, permeation, and release amounts from the breeding blanket system into the environment. In this paper, we introduce the integrated system and components of THETA-FR, shedding light on various tritium behaviors within the HCCP breeding blanket FW/BU component. By leveraging THETA-FR, researchers and engineers can make informed decisions regarding tritium control and safety in fusion reactors.

Deuterium retention in co-deposition with lithium in Magnum-PSI: experimental analysis and comparison with SOLPS-ITER

The vapor-box, a liquid metal design for the divertor, utilizes lithium recirculation through evaporation and condensation. Safety concerns arise from Li-D/T formation and co-deposition on vapor-box walls and the first wall, affecting tritium retention. Additively manufactured tungsten capillary porous structure (CPS) samples with Li were exposed to high heat flux D plasmas in the linear plasma device Magnum-PSI, to study D retention in Li–D co-deposition dependence on substrate temperature (200 ∘ C–428 ∘ C) and distance from the plasma beam center (25–85 mm). The D:Li ratio was determined via in-situ ion beam diagnostics with simultaneously analyzed Nuclear Reaction Analysis and Elastic Backscattering Spectroscopy spectra to maximize the precision. Experimental results approach close to the theoretical maximum at 40:60 D:Li ratio and deposited film thickness ranging from 0.02 to 3.2 µm. Witness plate temperatures above 400 ∘ C yielded Li films under 150 nm in thickness with lower D:Li ratios (5:95 D:Li ratio). At this temperature LiD decomposition pressure is comparable with vessel pressure during plasma. SOLPS-ITER simulations narrowed CPS surface temperature to 650 ∘ C–700 ∘ C, indicating Li+ plasma dominance near the target surface. Redeposition ratio of lithium on the CPS surface was determined to be around 80 % , matching quartz crystal microbalance results. However, SOLPS-ITER simulations lacked accuracy in recreating observed Li and D deposition layers on WPs, improvements are needed to model plasmas with significant Li quantities. Extension of SOLPS-ITER simulations to include LiD molecules and enhance heat flux accuracy is crucial for better alignment with experimental data.

Tritium behavior in soil and mineral rock components used for plant cultivation

Immersion, percolation and tritium release experiments in peat and vermiculite soil samples were performed to analyze their behavior in this widely used medium for plant cultivation. Samples were immersed in tritiated water for 696 h and the isotope exchange capacity evaluated. A vertical flow regime was also considered with analysis for hydraulic conductivity to understand tritium mobility and therefore its availability. Peat soil showed a high tritium retention after percolation, but vermiculite seem to suppress its retention ability. The high moisture and organic content of peat enhanced its isotope exchange capacity. The falling head method was used to numerically evaluate the saturated hydraulic conductivity and outflow flux. Calculated isotope exchange capacity was 4.95×10-2 mol-T2O/g for peat and 3.38×10-2 mol-T2O/g for vermiculite. The tritium release experiment showed significant release of tritiated carbons in peat.

Tritium transport modelling in Europe: Current status, multi-physics integration and proposal of material property data

The capability to predict tritium retention and permeation by accurate and reliable tritium transport models for the breeding blanket design development is one of the most relevant objectives of the European Test Blanket Module (TBM) Programme. This also represents a key aspect for the development and licensing of next-future fusion systems and power plants, including the European DEMO. Tritium analyses can be performed using either system-level models to calculate global blanket and the related ancillary system performances or detailed 3D models for spatially limited areas and critical regions: the choice between these two modelling approaches depend on the objective of each analyse. When looking at the accuracy of tritium transport simulations, two main aspects have to be considered: the first one is the completeness of the description of the physical phenomena participating to the tritium transport dynamics; the second one is the reliability of the tritium transport material property values used in the simulation activities.In this context, this paper has the objective to carry out a critical analysis on the status of the integration of the most relevant multi-physics elements, providing recommendations on the approach to follow, with a particular focus on the system-level codes. Moreover, a set of data and correlations on tritium transport materials properties is proposed as a first step towards a material property database to be shared within the breeding blanket community.

Modeling of tungsten impurity transport and distribution in EAST based on multi-fluid and kinetic Monte Carlo simulations

In recent years, the materials of plasma facing components, such as divertor target plates, domes, and outer walls of tokamaks, such as ASDEX Upgrade, WEST, JET, EAST, and ITER, have been changed from carbon to tungsten because of its lower erosion and tritium retention rates. Impurities are produced by interactions between the plasma and the first wall. This study provides an investigation to simulate the transport and distribution of tungsten impurities in the edge plasma on EAST. The 2D multi-fluid edge plasma transport code SOLPS-ITER and 2D kinetic Monte Carlo impurity transport code DIVIMP were used in the simulations. The multi-fluid model in SOLPS-ITER and the kinetic Monte Carlo model in DIVIMP were employed to treat tungsten impurity ions. The 2D density contour distributions in the computational region and the 1D density radial profiles at the inner and outer midplanes of tungsten impurity particles with ionization states (W0–W+74) and the total tungsten particles with all charge states were obtained. With the heating power 1.5 MW and the line-averaged plasma density 2 × 1019 m−3, the results from SOLPS-ITER and DIVIMP show that the maximum density of tungsten ion with single ionization state is about 1014 m−3 and the total density of tungsten impurities with all charge states is about 1015 m−3 at the core boundary. To the best of our knowledge, the simulation results from SOLPS-ITER and DIVIMP are compared for the first time to benchmark SOLPS-ITER with the multi-fluid mode and DIVIMP with the kinetic model for tungsten impurity transport. The density distributions of tungsten impurities with different ionization states from SOLPS-ITER and DIVIMP are highly similar, and good agreement can be found under similar conditions involved in the calculation. From the comparison benchmark between SOLPS-ITER and DIVIMP for tungsten impurity transport, it can be concluded that the impurity transport approximation used by DIVIMP is good.

Toward a high-fidelity tritium transport modeling for retention and permeation experiments

Tritium Migration Analysis Program version 8 (TMAP8), the latest version of TMAP, was developed within the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE). Created at Idaho National Laboratory (INL), MOOSE is an open-source, dimension-agnostic, fully coupled, and fully implicit multiphysics platform featuring massively parallel computation capabilities. Using TMAP8, tritium transport in a divertor monoblock was analyzed to elucidate the effects of pulsed operation (up to fifty 1,600 s plasma discharge and cool-down cycles) on the tritium in-vessel inventory source term and ex-vessel release term (i.e., tritium retention and permeation) for safety analysis. With its built-in Message Passing Interface capability, TMAP8 can, in under 2 h, simulate tritium transport in three different layered materials (i.e., tungsten, copper, and copper-chromium-zirconium alloy) in 2D geometry, using a single device/computer with 10 cores. The MOOSE-based TMAP8 code can leverage other MOOSE tools developed under the Nuclear Energy Advanced Modeling and Simulation program to perform tritium and thermal transport in complex geometries and multiphysics environments. And via its massively parallel computation, MOOSE will enable the fusion pilot plant designers to conduct high-fidelity multiphysics modeling for the design of the divertor and blanket systems as well as for the safety analysis.

Numerical Survey of the Sponge Mechanism for Reducing Tritium Retention in Fusion Reactors

Reducing the tritium inventory is one of the most important objectives in starting up a fusion reactor. To achieve this, the sponge mechanism, also known as deuterium presaturation, can be used to treat the surface of the tritium-contacting components prior to installation in the tritium recycling system. This treatment alleviates tritium from embedding in the material surface and reduces tritium retention in the reactor. Using a simplified physical model, the effect of the sponge mechanism on the minimum startup tritium inventory was simulated. Our results showed that the sponge effect can reduce the minimum startup tritium inventory by about 22%, shorten the tritium well time effectively, and extend the operational time determined by the tritium supply. Investigating the sponge mechanism is of great significance for designing a tritium recycling system.

An OpenFOAM multi-region solver for tritium transport modeling in fusion systems

The transport, permeation and retention of tritium inside nuclear fusion reactors is a topic of great interest due to the scarcity of the isotope, its ease of diffusion through materials and its radioactivity. An accurate balance of tritium is needed in all the fuel cycle, and each loss term needs to be evaluated in detail. In this context, reliable and flexible tools to evaluate tritium permeation and retention are a necessity for the development of fusion technologies. This work presents the OpenFOAM PermeAtion Solver for Tritium Analysis Foam (pastaFoam) solver and two custom boundary conditions, along with a series of verification and validation cases. The solver inherits all capabilities of the base solver, chtMultiRegionFoam, and is capable to simulate hydrogen transport in coupled fluid–solid systems in the presence of hydrogen traps, under diffusion limited regime or accounting for surface effects. All features are tested against analytical solutions and results are compared with other tritium transport codes. Agreement with experimental data is aligned with results of numerical benchmarks from literature.

Studies on the impacts of the pore size and shape on deuterium retention in tungsten fuzzy nanostructures

Tritium retention in plasma-facing materials is a critical issue that can significantly impact the long-term and steady-state operation of fusion devices. The experiments conducted in the laboratory device MIES have confirmed that the presence of the tungsten (W) nanostructure (called ‘fuzz’) leads to a substantial retention of hydrogen isotopes within W fuzz layer. This observation motivates us to conduct dedicated modeling to investigate the influence of W nanostructures on deuterium (D) retention using the three-dimensional kinetic Monte Carlo code SURO-FUZZ. The SURO-FUZZ code offers a great flexibility in generating diverse microscopic structures of the W fuzzy surface through the quartet structure generation set (QSGS) approach, which allows us to explore the effects of the pore size and shape on D retention. In this study, several different W nanostructures generated by QSGS approach are utilized to conduct a comprehensive comparison between MIES experiments and SURO-FUZZ simulations. It is demonstrated that the simulated D retention can be brought into a reasonable agreement with the experimental data. On this basis, predictive estimations of D retention on EAST and ITER have been performed with SURO-FUZZ modeling. The simulation results indicate that the total D retention induced by W fuzz remains well below the administrative limit of 700 g.

3D benchmark experiments of tritium in tungsten for tritium measurements by detecting β-ray induced X-rays

Tungsten is one of the most promising materials for plasma facing components (PFCs), and β-ray induced X-ray spectrometry (BIXS) is an important non-destructive method to obtain tritium depth profile and retention information for the development of PFCs, recovery of fuel and control of tritium safety in a fusion reactor. In this article, benchmark experiments to obtain tritium 3D profile in tungsten have been firstly performed using layer-by-layer chemical etching (LLCE) and imaging plate (IP) techniques. BIXS spectra have been detected at each layer. The average layer thickness of each erosion was controlled to be around 150 nm and 8 layers (the total thickness was 1.21 μm) were eroded, and the maximum uncertainty of tritium activity was 7.91 % in the LLCE experiments. Monte-Carlo simulations of BIXS spectra with Geant 4 code and four kinds of physical packages (Penelope, Livermore, EmStd. Opt1 and EmStd. Opt4) have been benchmarked against to the experimental results. Results show that all the four physical packages are not suitable to calculate the intensity of the BIXS spectra, and the current mode or cross sections should be improved to simulate the interaction between low electron/photon and tungsten accurately. However, the relative intensity of W-Mα peak and W-Lα can be simulated well using current physical packages such as EmStd. Opt4 and Livermore, which is a good indicator for tritium depth profile in tungsten.

Mechanism of hydrogen isotope exchange for tritium removal in plasma-facing materials: a multi-scale investigation

The safety of future fusion reactors is critically dependent on the tritium (T) retention in plasma-facing materials. Hydrogen isotope (HI) exchange offers a method to redistribute HIs within solid materials, presenting a feasible approach for removing T from bulk materials and trapped by strong trapping sites. Nonetheless, unraveling the intricate mechanism behind HI exchange remains an urgent yet formidable challenge. This study undertakes a comprehensive investigation into the mechanism of HI exchange in tungsten materials across multiple scales. First, we developed a multi-component hydrogen isotope transport and exchange model (HIDTX) based on classical rate theory. The model validation was further carried out, demonstrating good consistency with the well-controlled laboratory experiments. From the results of different comparative models in HIDTX, it is found that the reduction in deuterium retention due to HI exchange was primarily driven by three synergistic effects: competitive re-trapping, collision, and swapping effects. Through molecular dynamics (MD) and first-principles calculations, the microscopic mechanism of HI exchange was revealed to be that the presence of hydrogen atoms in the interstitial sites surrounding a vacancy in tungsten decreased the binding energy between the vacancy and hydrogen. Meanwhile, we discovered that the combination of thermal desorption and HI exchange can significantly lower the temperature required for the hydrogen removal and enhance the removal rate. Particularly, the hydrogen removal time can be shortened by approximately 95% with simultaneous HI exchange compared to that with only thermal desorption. This work provides a practical guideline for comprehending and subsequently designing for efficient T removal in future nuclear fusion materials.

Initial experiments to regenerate the surface of plasma-facing components by wire-based laser metal deposition

Plasma-facing components (PFC) in nuclear fusion reactors are exposed to demanding conditions during operation. The combination of thermal loads, plasma exposure as well as neutron induced damage and activation limits the number of materials suitable for this application. Due to its properties, tungsten (W) is foreseen as plasma-facing material (PFM) for the future DEMOnstration power plant. It is considered suitable due to its exceptionally high melting point, excellent thermal conductivity, low tritium retention and low erosion resistance during plasma exposure. But even tungsten armored PFCs have a limited lifetime due to, among other factors, surface erosion and the resulting thickness reduction of the armor material.In-situ local deposition of tungsten by means of additive manufacturing (AM) could counteract surface erosion and thus increase the service life span of PFCs. After evaluation of the potential AM processes qualified for this task, the wire-based laser metal deposition (LMD-w) process was selected as the most suitable process. First trials were conducted to examine if it is possible to reliably deposit tungsten onto tungsten substrate using the LMD-w process. In these first studies, single welding beads were generated, and in later experiments, entire layers were created from several welding beads which are arranged next to each other. To ensure reproducibility of the results, the substrate temperature was kept constant. Further experiments aimed at the elimination or minimization of problems such as oxidation, occurrence of balling defects, porosity, cracking, surface waviness and insufficient connection to the substrate. To increase the welding bead quality, the input parameters like laser power, deposition velocity, wire feed rate, inert gas flow, as well as the wire position were optimized. Furthermore, stacking of several layers, as well as the remelting of an already created layer, were carried out and investigated. This study represents the first steps in testing the feasibility of an in-situ surface regeneration concept for PFCs.

Investigation of gas and plasma driven deuterium permeation through a vanadium membrane for particle exhaust application in a fusion reactor

The process of particle exhaust based on the cryopump in fusion reactor presents a severe issue of deuterium and tritium retention. However, the metal foil permeation pump represents an effective means to solve the problem of deuterium and tritium retention during the particle exhaust process. How to enhance the permeability performance of metal films to meet the requirements of practical engineering applications of fusion reactor is the main content of the research. This article focuses on the simulation and experimental investigation of V membrane, aiming to explore methods for enhancing membrane permeation flux. Through simulation analysis, the influence of thickness, membrane surface state, and temperature on both plasma-driven and gas-driven permeation has been studied. Among these, the downstream surface condition is changed during the PDP process, resulting in an almost twofold increase in permeation flux. Based on the permeation experiments the influence of temperature on permeation was proved and the impact of plasma source power on permeation was also studied. It was found that the permeation flux increased more than two orders of magnitude when the radio frequency power changed from 0 to 500 W, which indicates that plasma incident flux serves as a more effective means to enhance permeation flux compared to temperature.

Overview of tritium retention in divertor tiles and dust particles from the JET tokamak with the ITER-like wall

Divertor tiles after Joint European Torus-ITER like wall (JET-ILW) campaigns and dust collected after JET-C and JET-ILW operation were examined by a set of complementary techniques (full combustion and radiography) to determine the total, specific and areal tritium activities, poloidal tritium distribution in the divertor and the presence of that isotope in individual dust particles. In the divertor tiles, the majority of tritium is detected in the surface region and, the areal activities in the ILW divertor are in the 0.5–12 kBq cm−2 range. The activity in the ILW dust is associated mainly with the presence of carbon particles being a legacy from the JET-C operation. The total tritium activities show significant differences between the JET operation with ILW and the earlier phase with the carbon wall (JET-C) indicating that tritium retention has been significantly decreased in the operation with ILW.

High-Temperature Intrinsic Defect Chemistry of Li8PbO6 Ceramic Breeding Material.

Understanding the intrinsic defect chemistry of tritium breeder materials proposed for use in future fusion reactors is imperative, as certain defects may act as traps leading to retention of tritium in the ceramic matrix. In this paper, we use combined density functional theory simulations with simple thermodynamics to explore the intrinsic defect chemistry of octalithium plumbate (Li8PbO6) as a function of both temperature and oxygen partial pressure. Importantly, we consider vibrational contributions to the energies of the reference states used in the calculations of the defect formation energies. Our results indicate that including these temperature effects can modify the predicted defect chemistry for materials at a high temperature. For Li8PbO6, the defect chemistry is predicted to be dominated by the VLi-1 defect, which will likely act as a trap for tritium. The charge compensating mechanism is predicted to change as a function of the conditions, with the Lii+1 interstitial defect providing compensation at low temperatures and the VO2+ vacancy defect occurring close to the Li2O saturation limit.

Readdressing nanocavity diffusion in tungsten

In nuclear fusion (ITER and the future DEMO), those components that face the plasma are exposed to high temperature and irradiation which, in the long term, modifies their thermal and mechanical properties and tritium retention. Tungsten is a candidate material and is the subject of many studies of microstructure evolution under various irradiation and temperature conditions. One milestone is the characterization of its defect properties. We here readdress the diffusion of nanocavities on broad ranges of size and temperature and compare it with dissociation, a competing process during nanocavity growth. First, at the atomic scale, we used molecular dynamics to explore the variety of elementary events involved in the nanocavity diffusion. Second, an experimental study of ion-irradiated samples, annealed at different temperatures up to 1,773 K, revealed the creation and growth of nanocavities on transmission electron microscopy images. Third, we performed multi-objective optimization of the nanocavity diffusion input of our object kinetic Monte Carlo model to reproduce the experimental results. Finally, we applied a sensitivity analysis of the main inputs of our model developed for these particular conditions—the source term which combines two cascade databases and the impurities whose interaction with the defects is characterised with a supplemented database of density functional theory calculations. Three domains of nanocavity size were observed. The first is the small vacancy clusters, for which atomistic calculations are possible and dissociation is negligible. The second is the small nanocavities, for which we provide new diffusion data and where a competition with the dissociation can take place. The third domain is the large nanocavities, for which, in any case, the dissociation prevents their existence above 1,500 K in the absence of a stabilizing interface.