Tritium Retention Research Articles

Oxidation resistance of bulk plasma-facing tungsten alloys

Tungsten (W) currently is the main candidate as plasma-facing armour material for the first wall of future fusion reactors, like DEMO. Advantages of W include a high melting point, high thermal conductivity, low tritium retention, and low erosion yield. However, in case of an accident, air ingress into the vacuum vessel can occur and the temperature of the first wall can reach 1200 K to 1450 K due to nuclear decay heat. In the absence of cooling, the temperature will remain in that range for several weeks. At these temperatures the radioactive tungsten oxide volatilizes. Therefore, ‘smart’ W alloys are developed that aim to preserve the properties of W during plasma operation coupled with suppressed tungsten oxide formation in case of an accident.This study focusses on oxidation studies at 1273 K of samples produced by mechanical alloying followed by field assisted sintering. In a first step the sintering is optimized for tungsten (W) – chromium (Cr) -yttrium (Y) alloys. It is shown that the best oxidation resistance is achieved with submicron grain sizes. This yields a closed, protective oxide layer. In a second step the influence of the grinding process during sample preparation is analysed. It is shown that scratches initiate failure of the protective oxide. In a third step the oxidation and sublimation is measured for weeks – for the first time the sublimation is directly measured in order to determine the potential hazard in comparison to pure W. It is shown that the oxidation is suppressed in comparison to pure W. However, sublimation at a rate of 1×10−6mgcm−2s−1 starts after a few days. Nevertheless, the progess in smart alloys is evident: sublimation is delayed by about two days and complete mechanical destruction of the first wall is avoided.

Recent accomplishments of the fusion safety program at the Idaho National Laboratory

The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences’ (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL.

Effects of baking in deuterium atmosphere on tritium removal from tungsten

Tritium-containing polycrystalline tungsten, which was pre-irradiated with helium prior to tritium exposure, was isothermally heated under a vacuum or deuterium gas in order to investigate the effect of the deuterium on tritium removal from the tungsten. It is found that in the case of 3 h of baking at temperatures varying from 423 K and 573 K, tritium retention after baking in deuterium gas became smaller than that in a vacuum. At the baking temperature of 573 K, tritium removal was hastened in the case of baking in deuterium atmosphere. The results indicate that the baking in deuterium atmosphere is effective to remove tritium from tungsten. The acceleration of diffusion of tritium in bulk and recombination of tritium with deuterium at the surface might be possible mechanisms.

Temperature dependence of liquid lithium film formation and deuterium retention on hot W samples studied by LID-QMS. Implications for future fusion reactors

Liquid metal (LM) divertor concepts explore an alternative solution to the challenging power/particle exhaust issues in future magnetic fusion reactors. Among them, lithium (Li) is the most promising material. Its use has shown important advantages in terms of improved H-mode plasma confinement and heat handling capabilities. In such scenario, a possible combination of tungsten (W) on the first wall and liquid Li on the divertor could be an acceptable solution, but several issues related to material compatibility remain open. In particular, the co-deposition of Li and hydrogen isotopes on W components could increase the associated tritium retention and represent a safety risk, especially if these co-deposits can uncontrollably grow in remote/plasma shadowed zones of the first wall. In this work, the retention of Li and deuterium (D) on tungsten at different surface temperature (200 °C–400 °C) has been studied by exposing W samples to Li evaporation under several D2 gaseous environments. Deuterium retention in the W–Li films has been quantified by using laser induced desorption-mass spectrometry (LID-QMS). Additional techniques as thermal desorption spectroscopy, secondary ion mass spectrometry, profilemetry and flame atomic emission spectroscopy were implemented to corroborate the retention results and for the qualitative and quantitative characterization of the films. The results showed a negligible (below LID sensibility) D uptake at Tsurface = 225 °C, when the W–Li layer is exposed to simultaneous Li evaporation and D2 gas exposition (0.67 Pa). Pre-lithiated samples were also exposed to higher D2 pressures (133.3 Pa) at different temperatures (200 °C–400 °C). A non-linear drastic reduction in the D retention with increasing temperatures was found on the W–Li films, presenting a D/Li atomic ratio at 400 °C lower than 0.1 at.% on a thin film of ≈100 nm thick. These results bode well (in terms of tritium inventory) for the potential utilization of this material combination in a real reactor scenario.

Estimation of the tritium retention in ITER tungsten divertor target using macroscopic rate equations simulations

Based on macroscopic rate equation simulations of tritium migration in an actively cooled tungsten (W) plasma facing component (PFC) using the code MHIMS (migration of hydrogen isotopes in metals), an estimation has been made of the tritium retention in ITER W divertor target during a non-uniform exponential distribution of particle fluxes. Two grades of materials are considered to be exposed to tritium ions: an undamaged W and a damaged W exposed to fast fusion neutrons. Due to strong temperature gradient in the PFC, Soret effect’s impacts on tritium retention is also evaluated for both cases. Thanks to the simulation, the evolutions of the tritium retention and the tritium migration depth are obtained as a function of the implanted flux and the number of cycles. From these evolutions, extrapolation laws are built to estimate the number of cycles needed for tritium to permeate from the implantation zone to the cooled surface and to quantify the corresponding retention of tritium throughout the W PFC.

A parametric study of helium retention in beryllium and its effect on deuterium retention

Beryllium samples have been exposed in the PISCES-B linear plasma device to conditions relevant to the International Thermonuclear Experimental Reactor (ITER) in pure He, D, and D/He mixed plasmas. Except at intermediate sample exposure temperatures (573–673 K) He addition to a D plasma is found to have a beneficial effect as it reduces the D retention in Be (up to ∼55%), although the mechanism is unclear. Retention of He is typically around 1020–1021 He m−2, and is affected primarily by the Be surface temperature during exposition, by the ion fluence at <500 K exposure, but not by the ion impact energy at 573 K. Contamination of the Be surface with high-Z elements from the mask of the sample holder in pure He plasmas is also observed under certain conditions, and leads to unexpectedly large He retention values, as well as changes in the surface morphology. An estimation of the tritium retention in the Be first wall of ITER is provided, being sufficiently low to allow a safe operation of ITER.

Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011–2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

Accumulation and Retention of Tritium (Tritiated Water) in Larval Fish (Carassius gibeliof and Radio Toxicological Effect

Accumulation and Retention of Tritium (Tritiated Water) in Larval Fish (Carassius gibeliof and Radio Toxicological Effect

The detection of He in tungsten following ion implantation by laser-induced breakdown spectroscopy

Abstract Laser-induced breakdown spectroscopy (LIBS) results are presented that provide depth-resolved identification of He implanted in polycrystalline tungsten (PC-W) targets by a 200 keV He+ ion beam, with a surface temperature of approximately 900 °C and a peak fluence of 10 23 m −2 . He retention, and the influence of He on deuterium and tritium recycling, permeation, and retention in PC-W plasma facing components are important questions for the divertor and plasma facing components in a fusion reactor, yet are difficult to quantify. The purpose of this work is to demonstrate the ability of LIBS to identify helium in tungsten; to investigate the sensitivity of laser parameters including, laser energy and gate delay, that directly influence the sensitivity and depth resolution of LIBS; and to perform a proof-of-principle experiment using LIBS to measure relative He intensities as a function of depth. The results presented demonstrate the potential not only to identify helium but also to develop a methodology to quantify gaseous impurity concentration in PC-W as a function of depth.

Background processes in the KATRIN main spectrometer

The KArlsruhe TRItium Neutrino (KATRIN) experiment is a large-scale experiment which aims for the model-independent determination of the effective mass of electron anti-neutrinos with a sensitivity of 200 meV/c2. It investigates the kinematics of electrons from tritium β-decay close to the endpoint of the energy spectrum. Low statistics at the endpoint requires an equally low background rate below 0.01 counts per second. The measurement setup consists of a high luminosity windowless gaseous molecular tritium source (WGTS), a differential and cryogenic pumped electron transport and tritium retention section, a tandem spectrometer section (pre-spectrometer and main spectrometer) for energy analysis, followed by a detector system for counting transmitted beta decay electrons. The background characteristics of the KATRIN main spectrometer were investigated in detail during two commissioning measurement phases. Of particular interest were backgrounds due to the decay of radon in the volume of the spectrometer, cosmic-muon-induced backgrounds, backgrounds due to natural radioactivity and Penning-discharge-related backgrounds. This proceeding will present results of the commissioning measurements and focuses on different background processes and their contribution to the overall background of the KATRIN experiment.

Hydrogen Isotope Retention and Permeation in Neutron-Irradiated Tungsten and Tungsten Alloys Under PHENIX Collaboration

Irradiation effects on heat-load and heat removal, thermo-mechanical properties, and tritium behavior in neutron-irradiated tungsten and tungsten alloy are being investigated under US-Japan PHENIX (Plasma facing components evaluation by tritium Plasma, HEat and Neutron Irradiation eXperiments) collaboration (2013–2018) to demonstrate feasibility and safety of helium-cooled divertor concept for a fusion demonstration (DEMO) and future fusion reactors. The PHENIX Task 3 is aimed at improved understanding of irradiation response on tritium retention and permeation in tungsten and tungsten alloys under divertor-relevant high-flux plasma for a Fusion Nuclear Science Facility (FNSF) and DEMO. This paper describes the challenge in elucidating tritium behavior in neutron-irradiated plasma facing components (PFCs), the PHENIX plans for neutron-irradiation and post irradiation examination, and progress in tritium behavior in neutron-irradiated tungsten.

Atomistic modeling of helium segregation to grain boundaries in tungsten and its effect on de-cohesion

Due to their low sputtering yield, low intrinsic tritium retention, high melting point, and high thermal conductivity, W and W alloys are promising candidates for the divertor region in a magnetic fusion device. Transmutation reactions under neutron irradiation lead to the formation of He and H particles that potentially degrade material performance and might lead to failure. High He fluxes ultimately lead to the formation and bursting of bubbles that induce swelling, a strong decrease in toughness, and a nanoscale microstructure that potentially degrades the plasma. Understanding the behavior of He in polycrystalline W is thus of significant importance as one avenue for controlling the material properties under operating conditions. This paper studies the interaction of substitutional He atoms with various grain boundaries in pure W and the effect of the He presence on the system response to external loading. We observe that He segregates to all the interfaces tested and decreases the cohesion of the system at the grain boundary. Upon tension, the presence of He significantly decreases the yield stress, which depends considerably on the bubble pressure. Increasing pressure reduces cohesion, as expected. More complex stress states result in more convoluted behavior, with He hindering grain boundary sliding upon simple shear.

Myth of initial loading tritium for DEMO—Modelling of fuel system and operation scenario

Any fusion power reactors such as “DEMO” that achieves tritium self-sufficiency with breeding blankets can produce tritium by DD reaction followed by exponential breeding in the blanket within reasonable total operation period. The present study further suggests that realistic Power Ascension Tests (PAT) of DEMO can produce its tritium to be needed in the series of tests by its own program until reaching steady state full power operation, with no external supply or additional operation costs. Closed tritium fuel plant was described by a system dynamics model, and analyzed considering realistic PATs of DEMO, that will be mainly pulsed DD and low concentration DT. Typical PATs require years of operation from zero power criticality to full power, with pulsed power output and long dwell time between them. Output power is gradually increased in PATs to check the functions of reactor systems and components. In the case of fusion DEMO, zero power criticality corresponds to DD operation. While plasma may be fired in pulses, tritium plant is continuously operated to recover all the tritium produced by the DD and low DT burn. Depending on the different time constant of tritium retention in components, tritium is transferred by deuterium purge, and high concentration tritium is finally collected in the storage, to be available for the next tests at virtually no additional cost. Realistic “initial loading tritium” will be 100g for commissioning of tritium plant far prior to the plasma operation.

Flowing liquid lithium plasma-facing components – Physics, technology and system analysis of the LiMIT system

The use of low atomic number liquid metals has been shown to have the potential to solve many of the prevalent problems like erosion and radiation losses associated with the interaction of fusion plasma with the plasma facing component (PFC) structures in tokamaks. Since the first evidence of lithium increasing plasma performance in TFTR [1], the benefits of using lithium in fusion environments have been seen in many devices, including CDX-U [2], NSTX [3], LTX [4], and DIII-D [5]. While both fast flow and slow flow concepts have been studied with regards to liquid lithium first wall alternatives, this report will focus on efforts placed on fast flow research and will mainly focus on advancements in the LiMIT device that help to eliminate concerns over the broad use of liquid lithium. Due to the promising TFTR results along with results obtained at the University of Illinois at Urbana-Champaign [6], suitably designed trench structures holding liquid lithium could be an appropriate fast flow candidate for PFC modules in future fusion devices. There are four potential shortcomings of this approach: (1) Droplet ejection, (2) Wetting control, (3) Tritium retention, and (4) Limited heat flux handling. Droplet ejection is discussed in a companion publication [7], while this paper addresses the topics of wetting control and heat flux handling. Limitations in wetting and prevention of lithium creep (i.e. getting and keeping the lithium only where it should be) have been solved by laser-texturing the base material with extreme short laser pulses (pico – femto second) of high power (several 10s of W). Micro- and nano-structuring results indicate that the textured substrates displayed significant change in their wetting properties, increasing the temperature needed to wet from 310 °C to 390 °C. Lastly, initial designs for the Lithium Metal Infused Trenches (LiMIT) [6] showed dryout above 3 MW/m2, but new designs of the trench shaping show potential to be able to handle up to 10MW/m2. Dryout is accompanied by lithium evaporation which is shown to mitigate the incident heat flux, which may be viewed as beneficial [8]. The advances shown here will increase the viability of the LiMIT system in large-scale testing, and allow for extensive design iteration to begin tackling the large powers and heat fluxes present in reactor-relevant systems.

Baseline high heat flux and plasma facing materials for fusion

In fusion reactors, surfaces of plasma facing components (PFCs) are exposed to high heat and particle flux. Tungsten and Copper alloys are primary candidates for plasma facing materials (PFMs) and coolant tube materials, respectively, mainly due to high thermal conductivity and, in the case of tungsten, its high melting point. In this paper, recent understandings and future issues on responses of tungsten and Cu alloys to fusion environments (high particle flux (including T and He), high heat flux, and high neutron doses) are reviewed. This review paper includes; Tritium retention in tungsten (K. Schmid and M. Balden), Impact of stationary and transient heat loads on tungsten (J.W. Coenen and Th. Loewenhoff), Helium effects on surface morphology of tungsten (Y. Ueda and A. Ito), Neutron radiation effects in tungsten (A. Hasegawa), and Copper and copper alloys development for high heat flux components (C. Hardie, M. Porton, and M. Gilbert).

Direct depth distribution measurement of deuterium in bulk tungsten exposed to high-flux plasma

Understanding tritium retention and permeation in plasma-facing components is critical for fusion safety and fuel cycle control. Glow discharge optical emission spectroscopy (GD-OES) is shown to be an effective tool to reveal the depth profile of deuterium in tungsten. Results confirm the detection of deuterium. A ∼46 μm depth profile revealed that the deuterium content decreased precipitously in the first 7 μm, and detectable amounts were observed to depths in excess of 20 μm. The large probing depth of GD-OES (up to 100s of μm) enables studies not previously accessible to the more conventional techniques for investigating deuterium retention. Of particular applicability is the use of GD-OES to measure the depth profile for experiments where high deuterium concentration in the bulk material is expected: deuterium retention in neutron irradiated materials, and ultra-high deuterium fluences in burning plasma environment.

Refractory Materials for Energy Applications

AbstractThe most extensively used refractory metals are tungsten, tantalum, niobium, rhenium, unalloyed molybdenum and its principal alloy, TZM. All of them are characterised by their extremely high melting point above 2000 °C (2273 K) and high hardness at room temperature. They are used in demanding applications that require high‐temperature strength and corrosion resistance, for example, wire filaments, casting moulds and chemical reaction vessels in corrosive environments. This contribution is focused on tungsten and new tungsten materials. The intrinsic problems of these materials, that is, their brittleness (even at elevated temperatures), thermal load capacity, heat conductivity, recrystallization, irradiation damage and specific defect processes, such as helium bubbles, swelling, new kinds of surface reactions, crack formation, brittle‐to‐ductile transition and tritium retention, require the development and assessment of new tungsten materials, and also new fabrication routes and processes.

Recent Upgrades at the Safety and Tritium Applied Research Facility

This paper gives a brief overview of the Safety and Tritium Applied Research (STAR) facility operated by the Fusion Safety Program (FSP) at the Idaho National Laboratory (INL). FSP researchers use the STAR facility to carry out experiments in tritium permeation and retention in various fusion materials, including wall armor tile materials. FSP researchers also perform other experimentation as well to support safety assessment in fusion development. This lab, in its present two-building configuration, has been in operation for over ten years. The main experiments at STAR are briefly described. This paper discusses recent work to enhance personnel safety at the facility. The STAR facility is a Department of Energy less than hazard category 3 facility; the personnel safety approach calls for ventilation and tritium monitoring for radiation protection. The tritium areas of STAR have about 4 to 12 air changes per hour, with air flow being once through and then routed to the facility vent stack. Additional radiation monitoring has been installed to read the laboratory room air where experiments with tritium are conducted. These ion chambers and bubblers are used to verify that no significant tritium concentrations are present in the experiment rooms. Standby electrical power has been added to the facility exhaust blower so that proper ventilation will now operate during commercial power outages as well as the real-time tritium air monitors.

Development and analyses of self-passivating tungsten alloys for DEMO accidental conditions

Tungsten is considered the main candidate material for the first-wall in DEMO due to its high melting point, low erosion yield and low tritium retention. Nevertheless, it can cause a substantial safety issue in a loss-of-coolant accident (LOCA) in combination with air ingress into the plasma vessel, due to the formation and sublimation of volatile neutron activated tungsten oxide. Self-passivating tungsten alloys introduce a passive safety mechanism by forming a stable chromic oxide scale on the surface acting as a diffusion barrier for oxygen and preventing the formation of tungsten oxide. Self-passivating tungsten alloys optimized for oxidation resistance containing ∼12wt.% Cr and ∼0.6wt.% Y are investigated under conditions of argon–oxygen, humid argon and humid air atmospheres at different partial pressures and temperatures ranging from 1073 to 1273K. Thin films with ∼3.5μm thickness produced by magnetron sputter deposition are used as a model system. The oxidation resistance of these films in an argon–20vol.% oxygen atmosphere is sufficient to prevent formation and release of tungsten oxide at temperatures of from 1073 to 1273K. The sublimation of Cr in nitrogen–oxygen–water atmosphere at T≥1273K is discussed. A deeper understanding of the governing processes for oxygen/Cr diffusion under different atmospheres is gained, supported by SEM/EDX in combination with FIB cross-section and TGA measurements.

Analysis and experimental verification of neutron-activated LiF as a stable matrix for tritium beta sources

Tritium (3H) has long been considered a useful radioisotope for many purposes due to its low-energy and pure beta decay. However, due to its aggressive migration, solid-state retention of tritium for use as a beta radiation source is challenging. Bulk-level infusion of tritium in a solid is often employed as a solution to effusion and outgassing but limits the beta radiation output due to the use of high-Z materials. This letter presents experimental and simulated analysis regarding the production and efficacy of a low-Z, solid-state tritium beta source. Single-crystal lithium fluoride (LiF) is used as a host matrix for tritium radionuclides, which are generated by the 6Li(n,α)3H reaction when the crystal is exposed to thermal neutron irradiation. The experimental findings present negligible outgassing of tritium from the material, while the simulation results suggest thickness optimization of the LiF/3H beta source. Moreover, the simulation results indicate significantly enhanced beta output efficiency compared to palladium tritide, which is a state-of-the-art tritium host. With proper thickness, the tritium-filled LiF matrix is a stable low energy beta source, which can be easily produced and incorporated into a variety of applications.