Wall Lifetime Research Articles

Time-dependent collisional radiative modeling of tungsten in the magnetic sheath for erosion diagnosis

Abstract Tungsten is the material of choice for the divertors in ITER, SPARC and future fusion reactors. Accurate diagnosis of tungsten erosion and migration is important for first wall life time, slag production and core performance. The addition of a magnetic presheath requires time-dependent collisional radiative effects to be included for accurate neutral tungsten collisional radiative modeling. Gross erosion measurements could be modified by a factor of 10 due to the inclusion of time-dependent effects for ITER relevant divertor conditions. A simple sputtering model and sheath density model are developed to investigate time-dependent collisional radiative effects. Neutral tungsten spectral lines populated from different metastable levels depend on model parameters leading to potential spectroscopic diagnostics of plasma parameters. Electron temperatures inferred from spectroscopic line ratios are in agreement with Langmuir probe measurements in the Compact Toroidal Hybrid.

Laser induced ablation spectroscopy for in situ characterization of the first wall on EAST tokamak

Plasma-wall interaction (PWI) is a crucial issue for high-performance operation and first wall lifetime of future fusion reactors. An in situ measurement and the control of fuel retention in the vacuum-vessel will be an important aspect for the safe operation of nuclear fusion devices. Laser-Induced Ablation Spectroscopy (LIAS) is an attractive option with promising capabilities for in situ characterization of fuel retention and co-deposition during plasma discharges. In this work, an in situ LIAS wall diagnostic system is developed for real-time during discharge diagnosis of the first wall on the EAST superconducting tokamak. A dedicated timing system, which offers the function of multiple temporal settings with high time resolution for LIAS measurement during long-pulse discharges, is successfully commissioned. The experimental setup of the LIAS system for the characterization of the first wall on EAST is presented in detail. First LIAS results obtained in EAST during high power discharges are presented. The results indicate that in situ measurements of deposition, co-deposition and dynamic retention are possible.

Improved pseudo-ductile behavior of powder metallurgical tungsten short fiber-reinforced tungsten (Wf/W)

For the first wall of a fusion reactor unique challenges on materials in extreme environments require advanced features in areas ranging from mechanical strength to thermal properties. The main challenges include wall lifetime, erosion, fuel management and overall safety. For the lifetime of the wall material, considerations of thermal fatigue due to transient heat loading are crucial as severe mechanical and thermal loads during operation are expected.Tungsten (W) is the main candidate material for the first wall of a fusion reactor as it is resilient against erosion, has the highest melting point of any metal and shows rather benign transmutation behavior under neutron irradiation. However, Tungsten has an issue related to intrinsic brittleness as well as operational embrittlement. To overcome this, a W-fiber enhanced W-composite material (Wf/W) incorporating extrinsic toughening mechanisms has been developed. Recently progress has been made in the powder metallurgical routes towards fully dense multi short-fiber Wf/W. For reasonable performance with respect to mechanical properties and hydrogen retention a fully dense pseudo-ductile Wf/W with is crucial. The properties of the used fibres are crucial. For the composite mechanisms to work a level of strength of the used fibres is required. In this contribution the change in ductility of the fibres is studied.In this contribution it is shown that excluding or minimising the impact of carbon impurities during the sintering process can significantly improve the mechanical properties of the fibres. New test results on the behaviour of PM Wf/W with and without a diffusion barrier during the sintering show a clear benefit as the fibres can retain ductility. Not the grain growth during sintering but the carbon present during sintering is clearly identified as determining the mechanical properties of the fibres.

Determination of Be sputtering yields from spectroscopic observations at the JET ITER-like wall based on three-dimensional ERO modelling

Estimations of the ITER first wall (FW) lifetime, previously made using the three-dimensional Monte-Carlo ERO code (Borodin et al 2011 Phys. Scr. T145 014008), depend strongly on the assumptions of the physical sputtering yield for beryllium (Be). It is of importance to validate the respective model and data at existing devices including the JET ITER-like wall (ILW) as most ITER-relevant experiments. Applying the same sputtering input data in ERO as those used before in the ITER-predictions, the ERO simulations for the Be light intensity (using up to date atomic data from ADAS and measured plasma conditions) reveal a factor of 2 overestimation in the assumed yield even if the low estimate assuming 50% D surface content is used. This result indicates the preference of this assumption for plasma-wetted areas. It points to a possible necessity to correct (reduce) the respective estimates for the Be sputtering yield and, accordingly, re-visit the ITER FW lifetime predictions.

Interior and Near-Wall Ion Velocity Distribution Functions in the H6 Hall Thruster

This paper presents the results of a two-axis laser-induced fluorescence velocimetry study of singly charged xenon in the interior and near the channel walls of the H6 Hall thruster. The thruster has a nominal operating discharge power of 6 kW and has been operated for before testing. It is operated under seven conditions spanning discharge voltages of 150–600 V and anode mass flow rates of . The mean velocity of the near-wall ion populations varies from 0.5 to with an angle of incidence relative to the wall surface that varies from 30 to 80 deg. In general, the mean velocity and angle of incidence are strongly correlated with the discharge voltage and weakly correlated with the anode mass flow rate. There is strong correlation between the axial locations where high energy (greater than a few tens of electron volts) ions exist and where erosion is discernible in the channel profile. This evidence supports the notion that erosion is negligible below a certain energy threshold. Potential application of this velocimetry technique for qualitative analysis of Hall thruster channel wall lifetime is also discussed.

Kinetic particle simulation of discharge and wall erosion of a Hall thruster

The primary lifetime limiting factor of Hall thrusters is the wall erosion caused by the ion induced sputtering, which is predominated by dielectric wall sheath and pre-sheath. However, so far only fluid or hybrid simulation models were applied to wall erosion and lifetime studies in which this non-quasi-neutral and non-equilibrium area cannot be treated directly. Thus, in this study, a 2D fully kinetic particle-in-cell model was presented for Hall thruster discharge and lifetime simulation. Because the fully kinetic lifetime simulation was yet to be achieved so far due to the high computational cost, the semi-implicit field solver and the technique of mass ratio manipulation was employed to accelerate the computation. However, other artificial manipulations like permittivity or geometry scaling were not used in order to avoid unrecoverable change of physics. Additionally, a new physics recovering model for the mass ratio was presented for better preservation of electron mobility at the weakly magnetically confined plasma region. The validity of the presented model was examined by various parametric studies, and the thrust performance and wall erosion rate of a laboratory model magnetic layer type Hall thruster was modeled for different operation conditions. The simulation results successfully reproduced the measurement results with typically less than 10% discrepancy without tuning any numerical parameters. It is also shown that the computational cost was reduced to the level that the Hall thruster fully kinetic lifetime simulation is feasible.

Erosion and Deposition Mechanisms in Fusion Plasmas

An overview of erosion and deposition processes in fusion machines is presented. The underlying physical and chemical mechanisms are explained. The impact of erosion and deposition on wall lifetime and tritium retention, which define the availability of future fusion machines such as ITER, is discussed. Also, examples of erosion and deposition observed in present fusion experiments are presented.

Measurements of beryllium sputtering yields at JET

The lifetime of the beryllium first wall in ITER will depend on erosion and redeposition processes. The physical sputtering yields for beryllium (both deuterium on beryllium (Be) and Be on Be) are of crucial importance since they drive the erosion process. Literature values of experimental sputtering yields show an order of magnitude variation so predictive modelling of ITER wall lifetimes has large uncertainty. We have reviewed the old beryllium yield experiments on JET and used current beryllium atomic data to produce revised beryllium sputtering yields. These experimental measurements have been compared with a simple physical sputtering model based on TRIM.SP beryllium yield data. Fair agreement is seen for beryllium yields from a clean beryllium limiter. However the yield on a beryllium divertor tile (with C/Be co-deposits) shows poor agreement at low electron temperatures indicating that the effect of the higher sputtering threshold for beryllium carbide is important.

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k eff ≤ 0.97 and produces less fission power of 1.5 GW th. Its main performance parameters turn out substantially better.

Erosion and Deposition Mechanisms in Fusion Plasmas

An overview of erosion and deposition processes in fusion machines is presented. The underlying physical and chemical mechanisms are explained. The impact of erosion and deposition on wall lifetime and tritium retention, which define the availability of future fusion machines such as ITER, is discussed. Also, examples of erosion and deposition observed in present fusion experiments are presented.

Contribution to DEMO reactor RH maintenance assessment

The scope of this paper is a preliminary assessment of the maintenance scheme in support of the European study for the next generation of fusion reactor: DEMO. Despite other fusion machine requiring remote handling maintenance operations, DEMO is supposed to work under steady state operational conditions. Therefore, requirement on the maintenance scheme is stronger. To target a good availability of the machine along machine operation plan, it is necessary to draw an adequate maintenance scheme. Indeed, due to the high fluxes generated by the plasma in the vacuum vessel, the first wall lifetime is limited, so the frequent replacement is necessary. On current fusion experimental machine, as first wall load conditions are less severe, DEMO condition implies high level of requirement on maintenance time. During DEMO lifetime, several full first wall replacements are expected. To provide access to the vacuum vessel machine for first wall removal, preparatory work is required to set the machine to adequate maintenance conditions and to open the machine properly, the same situation at the end of the maintenance period. Shutdown duration for first wall replacement should be as short as possible to reach the availability target of the machine. From this statement, the maintenance duration should not exceed 20% of the total lifetime of the reactor operation. First wall segmentation (i.e. total number of component to replace) has a high impact onto the replacement time. Considering the number of feasible designs for the first wall segmentation, we concentrate remote handling concept assessments one type of segmentation, the one minimizing the numbers of modules to replace [ITER technical basis; G A0 FDR 1 01-07-13 R1.0, Plant Description Document, Chapter 2.9; Remote handling at http://www.iter.org; ITER Conceptual Design Report, IAEA/ITER/DS/18, IAEA, Vienna, 1991; ITER Final Design Report, G A0 FDR 4 01-06-28 R 0.2, July 2001]. Assumption on Divertor segmentation for these DEMO studies have similarities with Divertor ITER design; therefore ITER design output is relevant [G. Janeschitz, A. Antipenkov, G. Federici, C. Ibbott, A. Kukushkin, P. Ladd, E. Mardin, R. Tivey, Divertor design and its integration into ITER, Nucl. Fusion 42 (2002), pp. 14–20; D. Maisonnier, Fusion Reactor Maintenance: A Proposal for the Power Plant Conceptual Study (PPCS), EFDA CSU Garching, N/R/0824/21/A,0.5.06.2000]. We assume divertor removal performed in shadow time, while removing the other first wall modules.

Lifetime performance analysis of HCPB power plant in-vessel components using HERCULES

The PPCS plant models explore a range of physics and technology assumptions spanning from near-term, ITER-like assumptions to very advanced scenarios. Plant model B (PMB) employs a helium-cooled, pebble bed (HCPB) blanket concept. Creep, both thermal and irradiation, is the primary limiting mechanism for in-vessel component lifetime, with the bulk of cumulative damage due to the power-on period. Lifetime improvement is possible if the time-to-failure at the full power-on stress level can be extended, which can be achieved through geometrical shape optimization. Redesign of the in-vessel build could rely on a slimmer tritium generating blanket (∼40 cm) instead of ∼50 cm, with the first wall acting as a “shielding” blanket ∼10 cm thick. The concept relies on extending the first wall and breeding blanket lifetime significantly, by limiting creep and fatigue, without losing the breeding capability of tritium breeding ratio (TBR) > 1. HERCULES was used for a neutronics analysis and the extraction of engineering parameters. A comparison with previous creep–fatigue analyses of helium-cooled blankets and first walls show that this new concept is capable of substantially increased lifetime, with TBR ∼ 1.0 (original TBR ∼ 1.14). The concept needs further refinement and optimization, but it is promising to deliver not only engineering parameters but also economic performance.

Erosion and Deposition Mechanisms in Fusion Plasmas

This paper gives an overview about erosion and deposition processes in fusion machines. The underlying physical and chemical mechanisms are explained. The impact of erosion and deposition on wall lifetime and tritium retention, which define the availability of future fusion machines such as ITER, is discussed.

Neutronic performance of HYLIFE-II fusion reactor using various thorium molten salts

HYLIFE-II is one of the major inertial fusion energy reactor design concepts in which a thick molten salt layer (Flibe = Li 2BeF 4) is injected between the reaction chamber walls and the explosions. Molten salt coolant eliminates the frequent replacement of solid first wall structure during reactors lifetime by decreasing intense neutron flux. This study presents the neutronic analysis of HYLIFE-II fusion reactor using various liquid wall coolants, namely, 75% LiF–25% ThF 4, 75% LiF–24% ThF 4–1% 233UF 4 or 75% LiF–23% ThF 4–2% 233UF 4. Neutron transport calculations for the evaluation of neutron spectra were conducted with the help of Scale 4.3 by solving the Boltzmann transport equation in S 8–P 3 approximation. The effects of flowing liquid wall thickness and type of coolant on the neutronic performance of the reactor were investigated. Furthermore, radiation damage calculations at the first wall structure with respect to type and thickness of liquid wall were carried out. Numerical results showed that using the flowing liquid wall containing the molten salt, 75% LiF–23% ThF 4–2% UF 4 with a thickness of ∼70 cm maintained tritium self-sufficiency of the (DT) fusion driver and extended the first wall lifetime to the reactors lifetime (∼30 full power years). In addition significant amount of high quality fissile fuel was bred through (n, γ) reaction of 232Th. Moreover, energy multiplication factor ( M) was increased to ∼12 by high rate fission reactions of 233U occurring in the flowing wall. On the other hand, it was concluded that using the other two coolants, 75% LiF–25% ThF 4 or 75% LiF–24% ThF 4–1% 233UF 4, as liquid wall did not satisfy the radiation damage and the tritium sufficiency criteria together at any thickness, so that these two coolants were not suitable to improve neutronic performance of HYLIFE-II reactor.

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF4 or ThF4 on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF4 whereas, 56 cm for that with Flibe+10% ThF4. Significant amount of high quality fissile fuel was produced by using heavy metal salt.

Impact of plasma, magnet and wall performances on tokamak and helical reactor economics

For the assessment of both tokamak and helical reactors, the base case reactor models are introduced, and parameter scan analyses on plasma beta, electric power output, maximum magnetic field strength and neutron wall loading are carried out. The assessment shows that high temperature operation is appropriate in tokamak reactors to increase bootstrap current fraction and current-drive (CD) efficiency. In contrast, low-temperature high-density operation is feasible and desirable in helical system to reduce helical ripple transport. The capital cost of helical reactors is rather high; however, the cost of electricity (COE) is not much higher than that of tokamak reactors especially in the case of low beta design, because of smaller re-circulation power (no current-drive power) and less-frequent blanket replacements due to lower neutron wall load. The engineering improvement of increasing maximum magnetic field strength gives rise to the compact designs with lower construction cost. However, it does not lead to the strong reduction in COE because of the increase in the neutron wall loading. The COE critically depends on the system availability which is determined by the wall lifetime and the blanket maintenance time.

Material erosion and migration in tokamaks

The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the susbsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasma-facing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined.

Overview of JET results

High density and high confinement operation in ELMy H-mode is confirmed at or above the normalized parameters foreseen for the ITER operating point (H98(y,2) ∼ 1, n/nGW ∼ 1, βN > 1.8 at q95 ∼ 3). The scaling of the ELMy H-mode with βN could be more favourable than that predicted by the IPB98(y,2) scaling. In ELMy H-mode, ion cyclotron current drive (ICCD) control of large sawteeth stabilized by fast particle has been demonstrated and the underlying neo-classical tearing modes (NTMs) and sawtooth physics is being refined. At high-density, Type I ELMy H-modes show trends that would lead to marginally acceptable ELMs on ITER. Type II ELM regime has been produced, though under very restrictive conditions. Type III ELMy operation with radiation fractions up to 95% has been demonstrated by seeding of N2 in H-modes and could extrapolate to Q = 10 ITER operation, albeit at high current (17 MA). The mitigation of Type I ELMs, nevertheless, remains a challenge. Considerable progress has been obtained in internal transport barrier (ITB) plasmas, with operation at central densities close to the Greenwald density or/and low toroidal rotation or/and high triangularity. Demonstrations of full current drive and successful simultaneous real time control of safety factor and temperature profiles have been achieved in ITB plasmas. Physics of resistive wall modes (RWMs) has been compared with theory, showing favourable scaling for ITER. High βN ∼ 2.8 operation of hybrid modes (also called improved H-modes) has been obtained with dominant neutral beam heating. Hybrid modes with dominant ion cyclotron resonance heating (ICRH) have also been achieved. Trace tritium experiments yielded valuable information on particle transport in H-mode, ITB and hybrid regimes. In Type I ELMy plasmas, successful tests of the conjugate-T ICRH scheme have been achieved as well as lower hybrid coupling at ITER-relevant 10–11 cm distances. Reduced D and T fuel retention has been observed, which could relate to operation with vertical targets in the divertor and/or lower (ITER-like) vessel temperature. It is confirmed that erosion occurs predominantly on the main chamber surfaces, with possible benefits for T retention in ITER, although consequences for the metallic first wall lifetime need to be assessed. Disruption and ELM studies indicate that transient power deposition could be less constraining than expected for the ITER divertor, but more challenging for the metallic first wall. Alpha particle tomography and direct observation of alpha particle slowing down have been made possible by γ-spectroscopy. Measurements of Alfvén cascades have been improved by a new interferometric technique. Promising tests of ITER relevant neutron counting detectors have been conducted.

New critical assessments of chamber and wall response to target implosion in inertial fusion reactors

The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to next target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, neutrons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive heights software package. The hydrodynamic response of gas-filled cavities and photon radiation transport of the deposited energy have been calculated by means of new and advanced numerical techniques for accurate shock treatment and propagation. Photon radiation transport models are developed for either the gas-filled cavity or in the evolving vapor cloud layer above the wall surface. The focus of this work is to examine the overall wall response and lifetime due to various erosion mechanisms.

Blister formation and erosion due to blister fracture of SiC or Si

For silicon wafer and silicon carbide, helium ion irradiation followed by annealing was repeated, and the weight loss was measured. The blister formation and its fracture processes were also examined. For Si, many blisters were observed at the surface after the first ion irradiation. After the annealing at 1173 K after the irradiation, small pores formed by helium desorption were observed. With the increase of the cycle number, surface roughness gradually increased. The weight loss due to annealing was about a half of that by ion sputtering. The sputtering yield due to the ion irradiation was approximately twice the reference value in the present experiment, perhaps due to the bubble fracture during the irradiation. For SiC, small blisters were observed on SiC crystals after the ion irradiation. The erosion amount due to only the process of annealing at 1323 K was several times larger than that of ion sputtering, although the sputtering yield due to the ion irradiation was roughly the same as the reference value. The present erosion was significantly large, so that this erosion has to be taken into account regarding the wall lifetime and impurity level of the core plasma.