Helium Retention Research Articles

Helium cluster dissolution in molybdenum

Helium retention and diffusion in molybdenum is studied on an atomistic scale withab initio methods. The thermal stability of helium–vacancy clusters is quantified within theframework of density functional theory. Calculated helium emission rates are used to derivea desorption spectrum which is compared with experimental results. The agreementbetween the current calculations and available experiments is satisfactory except in thehigh temperature end of the spectrum. The current results indicate that above 1100 K Hemigration is assisted by lattice defects such as vacancies, rather than through interstitialdiffusion.

Helium Retention and Desorption Behaviour of Reduced Activation Ferritic/Martenstic Steel

The reduced activation ferritic/martenstic steel CLF-1 prepared by the Southwestern Institute of Physics in China was irradiated by helium ions with an energy of 5 keV at room temperature using an electron cyclotron resonance (ECR) ion irradiation apparatus. After the irradiation, the helium retention and desorption were investigated using a technique of thermal desorption spectroscopy (TDS). The experiment was conducted with both the normal and welded samples. Blisters were observed after the helium ion irradiation, and the surface density of blisters in the welded samples was lower than that in the non-welded samples. Three desorption peaks were observed in both the non-welded and welded samples. These desorption peaks corresponded to those of blister ruptures and the helium release from the inner bubbles and the defects. The amount of helium retained in the welded samples was approximately the same as that in the non-welded samples, which was much less than other reduced activation materials, such as vanadium alloy and SiC/SiC composites.

Difference between helium retention properties in 316L and 304 stainless steels

The difference between helium behavior in type 316L and 304 stainless steels at high temperature was examined. Thermal desorption experiments were performed to obtain fundamental knowledge on the retention properties of helium atoms and in situ TEM observation was also carried out to examine the related dynamic behavior of helium bubbles. The desorption properties indicated that SUS316L has strong retention up to high temperatures. In SUS316L, a large desorption peak was present at temperatures above 1200 K, whereas most of the retained helium was desorbed at comparatively low temperatures in SUS304. In situ TEM observation consistently showed rather lower mobility of bubbles in SUS316L compared with SUS304. The suppression of bubble motion could be caused by Mo atoms, the additive in SUS316L.

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G∼100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ∼ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Reduction of hydrogen and helium retention in stainless steel by argon glow discharge

Although wall conditioning using Ar glow discharges is occasionally employed to reduce helium and hydrogen retention in the inner walls of the vacuum chamber of nuclear fusion devices, the effect of an Ar glow discharge on the reduction of helium and hydrogen retention has never been systematically investigated. In the present experiment, Ar glow discharges were applied to the stainless steel walls of a glow discharge device. Residual gas analysis (RGA) clearly showed that the Ar glow discharge significantly reduced the amount of helium and hydrogen retained. It was found that the fraction of the amount of desorbed helium to the initial retained amount of helium was as high as 13% and the fraction of the amount of desorbed hydrogen to the initial amount of hydrogen was as high as 36%.

Retention and desorption behavior of helium in oxidized V–4Cr–4Ti alloy

Retention and desorption behaviors of helium in oxidized and non-oxidized V–4Cr–4Ti alloy samples were investigated after helium ion irradiation at room temperature using a thermal desorption spectroscopy. The ion energy and fluence were 5 keV and (0.5–10) × 10 21 He/m 2, respectively. An oxidized layer with a thickness of 100 nm was prepared by thermal oxidation. The surface density of blisters produced by helium ion irradiation in the oxidized sample was lower than that in the non-oxidized one. The helium desorption behavior depended significantly on the fluence. In the lower fluence regime, the retained helium desorbed mainly at around 1300 K in both samples. As fluence increased, several desorption peaks appeared in the low temperature region in both samples. However, the peak temperatures were different. The amount of helium retained in the oxidized sample was lower than that in the non-oxidized sample.

Radiation damage and helium diffusion kinetics in apatite

Experimentally determined diffusion coefficients for 39 different samples of apatite demonstrate that the closure temperature (Tc) for helium retention in apatite spans a wider range than previously recognized: from 44 ± 4 C to 116 ± 18 C (for 10 C/ Myr) and correlates with the radiogenic He concentration ([He]) in a given sample. We find no correlation between helium diffusion kinetics and apatite chemistry, including the F/Cl ratio. We argue that [He] is a measurable proxy for the radiation damage which accumulated within each crystal over geologic time. As the volume density of lattice damage sites increases, apatite becomes more helium retentive. This implies that helium retentivity, and hence the effective helium diffusion kinetics, is an evolving function of time. Measured diffusivities thus reflect a snapshot in time and cannot alone be applied to the thermochronometric interpretation of a given sample. Calibrated with diffusion kinetics of 39 different samples of apatite, we present a simple, quantitative trapping model which relates diffusivity to both temperature and [He]. This previously proposed model (Farley, 2000) consists of two Arrhenius relations: one for volume diffusion through undamaged mineral lattice and one for the release of helium from the damage traps back into the undamaged lattice. The model predicts much of the observed log-linear correlation between Tc and [ He]. By inserting this function into a He production-diffusion calculation, the trapping model predicts: (i) that the effective He closure temperature of apatite will vary with cooling rate and effective U concentration (eU) and may differ from 70 C by up to ±15 C, (ii) the depth of the He partial retention zone will depend on accumulation time and on eU, and (iii) samples subjected to reheating after the accumulation of substantial radiation damage will be more retentive than previously expected. These predictions are consistent with recent observations of unexpected apatite (UTh)/He ages in some settings, most noteably (Flowers et al., 2006).

Helium thermal desorption and retention properties of V–4Cr–4Ti alloy used for first wall of breeding blanket

Helium irradiation experiments of V–4Cr–4Ti alloy with various surface treatments were conducted in an electron cyclotron resonance (ECR) ion irradiation apparatus. After helium ion irradiation at room temperature, the helium thermal desorption and retention properties were examined by thermal desorption spectroscopy (TDS). Ion energy of helium beam was 5 keV. Three groups of desorption peaks appeared at around 500, 850 and 1200 K in the TDS spectrum. After helium ion irradiation at ion fluence of 1 × 10 21 He/m 2, the retained helium desorbed mainly at around 1200 K and all of the implanted helium atoms were retained. With increasing fluence up to 5 × 10 21 He/m 2, the amount of helium desorbed at 500 K increased. For the polished samples with annealing at various temperatures, the desorption peak observed at around 500 K shifted to higher temperature region. Smallest retained amount of helium was observed in the V-alloy with polishing followed by annealing at 1373 K.

06/00098 Retention and desorption of hydrogen and helium in stainless steel wall by glow discharge: Hino, T. et al. Fusion Engineering and Design, 2005, 72, (4), 339–344

06/00098 Retention and desorption of hydrogen and helium in stainless steel wall by glow discharge: Hino, T. et al. Fusion Engineering and Design, 2005, 72, (4), 339–344

Deuterium retention of low activation ferritic steel and boronized wall in JFT-2M

Deuterium retention and erosion of F82H exposed to JFT-2M plasmas and the effect of boronization on the deuterium retention were examined based on material probe analysis. The amount of retained deuterium almost saturated after 300 tokamak discharges of deuterium. The retention of helium was clearly observed after helium glow discharge. The desorption temperature was approximately 600 K. While no significant change in the surface atomic composition was observed after the exposure of tokamak discharges, the mixture layer of carbon, iron and oxygen was observed after the helium glow discharge due to surface erosion and redeposition of eroded particles. After the boronization using trimethylboron, both carbon and boron deposited at the surface. The atomic ratio of B to C was 1/3. The amount of retained deuterium significantly increased, three times larger than that before the boronization, owing to that a large amount of deuterium is trapped in the carbon content.

Progress of plasma surface interaction study on low activation materials

Ferritic steel, vanadium alloy and SiC/SiC composite are candidate low activation materials for blanket components and first walls in fusion demonstration reactors. Several issues on these materials as the first wall have been investigated so far. Amount of deuterium retained in mechanically polished ferritic steel, F82H, after deuterium ion irradiation, was observed to be several times smaller than that of stainless steel, 316L SS. Physical sputtering yield of the ferritic steel due to deuterium ion was comparable to that of 316L SS. These results suggest that the property of the ferritic steel as the first wall material is superior to that of 316L SS, with respect to fuel hydrogen retention and in-vessel tritium inventory. Since first walls of blanket modules are exposed to both fuel hydrogen and helium, the helium is also trapped in the walls. Helium retention of V–4Cr–4Ti was investigated using helium ion irradiation apparatus. The amount of helium retained was comparable to those of other plasma facing materials. One of the major concerns in use of SiC/SiC composite for blanket is permeation of helium gas coolant into fusion plasma. Helium gas permeability of the SiC/SiC composite after heat cycles was measured using a vacuum device consisting two chambers. The increase in the permeability was not observed when the heating rate was suitably adjusted. Therefore, the blanket module may be made using only SiC/SiC composite if a vacuum pumping for the inside of blanket module is attached.

Retention and surface blistering of helium irradiated tungsten as a first wall material

The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion irradiation and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature, pulsed implantation, and tungsten microstructure was conducted to better understand what may occur at the first wall of the reactor. Single crystal and polycrystalline tungsten samples were implanted with 1.3 MeV 3He in doses ranging from 10 19 m −2 to 10 22 m −2. Implanted samples were analyzed by 3He(d,p) 4He nuclear reaction analysis and 3He(n,p)T neutron depth profiling techniques. Surface blistering was observed for doses greater than 10 21 He/m 2. For He fluences of 5 × 10 20 He/m 2, similar retention levels in both microstructures resulted without blistering. Implantation and flash heating in cycles indicated that helium retention was mitigated with decreasing He dose per cycle.

An overview of the development of the first wall and other principal components of a laser fusion power plant

This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0–100 keV) and ions (0.1–4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential candidate FW armor materials.

Microstructural analysis on helium retention of ion-irradiated and annealed tungsten foils

The helium retention characteristics and helium bubble distribution in tungsten were studied using 3He(d,p) 4He nuclear reaction analysis (NRA) and transmission electron microscopy (TEM) on two forms of tungsten: single crystal and polycrystalline, implanted to 1 × 10 19 3He/m 2 at 850 °C and annealed at 2000 °C. The NRA results revealed that as-implanted single crystal and polycrystalline tungsten exhibited similar helium retention characteristics. Stepwise annealing reduced the helium retention in both single crystal and polycrystalline tungsten when the number of implantation steps and annealing time were increased. The TEM results indicated that microstructure played a large role in helium trapping; the existence of grain boundaries led to significant cavity formation and greater cavity growth. Single crystal tungsten had less trapping sites for helium, allowing long range He diffusion during annealing. The decrease of He retention in polycrystalline tungsten during stepwise annealing was probably due to significant recrystallization, resulting in decrease of grain boundary density.

Deuterium and helium retentions of V–4Cr–4Ti alloy used as first wall of breeding blanket in a fusion reactor

The deuterium and helium retention properties of V–4Cr–4Ti alloy were investigated by thermal desorption spectroscopy (TDS). Ion energies of deuterium and helium were taken at 1.7 and 5 keV, respectively. The retained amount of deuterium in the sample irradiated at 380 K increased with the ion fluence and was not saturated to fluence of up to 1 × 10 23 D/m 2. For the irradiation at 773 K, 0.1% of implanted deuterium was retained at the highest fluence. For the helium ion irradiation at room temperature, three groups of desorption peaks appeared at around 500, 850, and 1200 K in the TDS spectrum. In the lower fluence region (<1 × 10 21 He/m 2), the retained helium desorbed mainly at around 1200 K. With increasing fluence, the amount desorbed at 500 K increased. Total amount of retained helium in the samples saturated at fluence up to 5 × 10 21 He/m 2 and saturation level was 2.7 × 10 21 He/m 2.

Effect of heat treatment on helium trapping in vanadium alloy at high ion implantation fluence

Helium was implanted into a V–4Ti alloy by using a 5keV helium ion beam with a flux of 1014He/cm2s, and then helium retention was analyzed by a technique of thermal desorption spectroscopy (TDS). The influence of heat treatment on helium trapping was studied by pre-annealing of the samples before ion irradiation. The annealing temperature was 973, 1223 and 1373K, respectively. Results indicated that pre-annealing treatments did not influence significantly the helium retention and release behavior in vanadium alloy on the present experimental conditions where the helium ion implantation fluence is in the region of 1017–1018 He/cm2, which shows a different tendency from the situation at a low fluence region (1013–1014 He/cm2). The reason could be due to different trap mechanisms of helium at different ion fluences.

Helium retention and surface blistering characteristics of tungsten with regard to first wall conditions in an inertial fusion energy reactor

The first wall of an inertial fusion energy reactor may suffer from surface blistering and exfoliation due to helium ion fluxes and extreme temperatures. Tungsten is a candidate for the first wall material. A study of helium retention and surface blistering with regard to helium dose, temperature and tungsten microstructure was conducted to learn how the damaging effects of helium may be diminished. Single crystal and polycrystalline tungsten samples were implanted with 1.3MeV 3He in doses ranging from 1019/m2 to 1022/m2. Implanted samples were analyzed by 3He(d,p)4He nuclear reaction analysis and neutron depth profiling techniques. Surface blistering occurred for doses greater than 1021He/m2 and was analyzed by scanning electron microscopy. Repeated cycles of implantation and flash annealing indicated that helium retention was reduced with decreasing implant dose per cycle. A carbon foil energy degrader, currently in development, will allow a continuous spectrum of helium implantation energy matching the theoretical models of He ion fluxes within the IFE reactor.

First principles study of the alloying effect on chemical bonding characteristics of helium in La–Ni–M tritides

The alloying effect on the electronic structure of La–Ni–M tritides is investigated using the first principles discrete variational Xα(DV-Xα) method. The calculated results show that the covalent interaction between atoms will play a much more important role in studying the alloying effect on chemical bonding characteristics in La–Ni–M tritides than ionic interaction. It is also found that in La–Ni–M tritides helium forms stronger covalent bonds with the weaker hydride forming elements than La. By analyzing the relation between the binding energy difference and bond order, our study indicates that after some alloying elements substituting for Ni locating in 3g site in tritides, the helium retention capability becomes stronger, changes as the following sequence: Al > Cr > Mn > Fe > Co > Ni, and is also very distinct for Cu although the chemical bonding between Cu atom and Ni atom is degraded drastically.

05/01531 Retention and desorption of hydrogen and helium in stainless steel wall by glow discharge

05/01531 Retention and desorption of hydrogen and helium in stainless steel wall by glow discharge