Development of Deuterium-Tritium Loading Vessel for Laser Fusion Target

Tritium permeation at 350°C through stainless steel wall of a vessel filled with deuterium-tritium gas of 6.1 × 10⁶ Pa pressure was practically suppressed by Au plating of 20μm thick applied to the outside surface. The apparent diffusivity of hydrogen through plated Au layer, derived from the experimental data, was 2 x 10^?11 cm²/s for 470°C, which is 10^?5–10^?6 times smaller than what would be expected from values reported for wrought Au, and the apparent solubility was very significantly higher than similarly expected level. Gas analysis of the Au layer indicated that the effective suppression of tritium permeation is attributable to trapping of hydrogen by C contained in the Au as impurity. Adequate tightness against tritium leakage has been achieved by Au plating on a vessel used for loading glass microspheres with deuterium-tritium gas, intended for laser fusion targets.     

Impact of nuclear irradiation on helium bubble nucleation at interfaces in liquid metals coupled to permeation through stainless steels

The impact of nucleating gas bubbles in the form of a dispersed gas phase on hydrogen isotope permeation at interfaces between liquid metals, like LLE, and structural materials, like stainless steel, has been studied. Liquid metal to structural material interfaces involving surfaces, may lower the nucleation barrier promoting bubble nucleation at active sites. Hence, hydrogen isotope absorption into gas bubbles modelling and control at interfaces may have a capital importance regarding design, operation and safety.He bubbles as a permeation barrier principle is analysed showing a significant impact on hydrogen isotope permeation, which may have a significant effect on liquid metal systems, e.g., tritium extraction systems. Liquid metals like LLE under nuclear irradiation in, e.g., breeding blankets of a nuclear fusion reactor would generate tritium which is to be extracted and recirculated as fuel. At the same time that tritium is bred, helium is also generated and may precipitate in the form of nano bubbles.Phenomena modelling is exposed and implemented in openFROM^® CFD tool for 0D to 3D simulations. Results for a 1D case show the impact of a He dispersed phase of nano bubbles on hydrogen isotopes permeation at an interface. In addition, a simple permeator simulation, consisting in a straight 3D pipe is exposed showing the effect of a He dispersed gas phase on hydrogen isotope permeation through different stainless steels. Results show the permeation reduction as a function of the interface area covered by He bubbles.Our work highlights the effect of gas bubble nucleation at interfaces and the importance of controlling these phenomena in nuclear technology applications.     

氚在真空贮存容器材料中的吸附和渗透行为

研究了贮存氚靶约4 a和20 a的两个316 L不锈钢真空贮存容器(以下简称贮存容器)及其垫片材料对氚的吸附行为,并对氚在贮存容器材料中的渗透速率进行了测量和分析。结果表明,贮存容器外表面氚污染为几十Bq/cm2,不锈钢与陶瓷中吸附的氚活度均为106Bq/g;热解吸至1 273 K过程中,材料中99%的氚释放出来;在解吸出的氚中,陶瓷中的HTO比例高于不锈钢;贮存温度对氚靶贮存容器的渗氚速率有较大影响,夏季约为冬季的4倍。上述结果提示,氚在贮存容器材料内表面吸附后,一部分会向晶格扩散并滞留下来;另一部分则透过材料向外环境渗透,其中温度是影响氚向外环境渗透的主要因素之一。     

Experimental study of efficiency of natural oxide layers for reduction of tritium permeation through Eurofer 97

Tritium permeation from the breeder through the helium coolant is a fundamental safety issue in the design of the HCLL (Helium coolant lithium lead) blanket system. The permeation of hydrogen isotopes through Eurofer in different conditions was deeply studied in the past, demonstrating that it is necessary to reduce this amount using tritium permeation barriers (TPB). A strong effort has been made to select the best technological solution for the realisation of tritium permeation barriers on complex structures not directly accessible after the completion of the manufacturing process, but after many years of activity for the qualification of different materials acting as TPBs it was demonstrated that these technologies are not yet mature for nuclear applications. An easier solution was identified in the nucleation and growth of natural oxides on the helium-exposed surface of cooling system components. The major objective of this work is the evaluation of the Permeation Reduction Factor (PRF) of natural oxides on Eurofer steel adding a known content of water and hydrogen to argon, used in substitution of helium. The PRF was measured on disk shaped specimens, in gas phase, using the PERI II apparatus, at a temperature of 550 °C. The oxide layer was produced in situ, in a well-defined range of hydrogen on water ratios. The obtained results are presented and discussed.     

Characterization and storage of radioactive zeolite waste

For the safe storage of zeolite wastes generated by the treatment of radioactive saline water at the Fukushima Daiichi Nuclear Power Station, this study investigated the fundamental properties of herschelite adsorbent and evaluated its adsorption vessel for hydrogen production and corrosion. The hydrogen produced by the herschelite sample is oxidized by radicals as it diffuses to the water surface and thus depends on the sample's water level and dissolved species. The hydrogen production rate of herschelite submerged in seawater or pure water may be evaluated by accounting for the water depth. From the obtained fundamental properties, the hydrogen concentration of a reference vessel (decay heat = 504 W) with or without residual pure water was evaluated by thermal–hydraulic analysis. The maximum hydrogen concentration was below the lower explosive limit (4%). The steady-state corrosion potential of a stainless steel 316L increased with the absorbed dose rate, but the increase was repressed in the presence of herschelite. The temperature and absorbed dose at the bottom of the 504 W vessel were determined as 60 °C and 750 Gy/h, respectively. Under these conditions, localized corrosion of a herschelite-contacted 316L vessel would not immediately occur at Cl^? concentrations of 20,000 ppm.     

Studies on the Characteristic of Titanium–Tritium Reaction

The p–c–T curves of tritium absorption and desorption of titanium were measured using the method of step equilibrium by stepping up the tritium quantity on an experimental apparatus of metal hydride. The p–c–T curves for tritium have one plateau at temperature below 300°C and two plateaus at temperature above 300°C. The thermodynamic parameters of the different phases were determined according to the van’t Hoff equation. The hysteresis effect was not observed in reversible process of tritium absorption and desorption of titanium on our experimental condition. The tritium absorption behavior of titanium in the temperature ranging from 550°C to 750°C and desorption behavior of titanium in the temperature ranging from 350°C to 550°C have been investigated in a constant volume system. A method of the reaction rate analysis was proposed and examined for determining the rate constant. The apparent activation energy obtained by this analysis for the absorption and the desorption were 155.5 ± 3.2 kJ mol⁻¹ and 62.1 ± 1.6 kJ mol⁻¹ respectively.     

Investigation of fusion reactions in palladium and titanium tritide using galvanostatic,coulometric, and hydrogen permeation techniques

We describe several electrochemical methods used to investigate the possibility of cold fusion phenomena in palladium and titanium tritide cathodes. We performed long-term (up to 77 days) electrolysis experiments with electrochemical cells of the University of Utah type at current densities as high as 1 A/cm², while monitoring neutron and tritium levels. With some cells, we pulsed the current to determine if neutron bursts would result. In another cell, we used titanium tritide as the cathode to determine if D-T reactions yielding neutrons would occur. In no instance were levels of neutrons or tritium significantly above background except in the titanium tritide cell where isotopic exchange, occcurring between the electrode and the electrolyte, resulted in significant tritium levels. We also combined x-ray photoelectron spectroscopy (XPS) and electrochemical hydrogen permeation experiments to determine the effectiveness of various Pd surface treatment procedures on the resultant electrochemical hydrogen absorption efficiency. Electroanalytical and thermal desorption/gas analysis techniques indicated the maximum loading of H in Pd was to a ratio of HPd=0.8.     

Surface mediated isotope exchange reactions between water and gaseous deuterium

Maintaining isotopic purity of hydrogen is one of the major tasks in tritium processing systems. The work with multiple isotopes and isotopomers is accompanied by isotope exchanges which is often accelerated by catalysts e.g. surfaces of various materials. In this work, densities of D₂O, HDO produced via isotope exchange reactions in the mixture of D₂, H₂, D₂O, H₂O, HD and HDO contained in a stainless steel (type SS304) vessel were measured as a function of time (40-36 000 s) and pressures near 3.5 × 10² Pa, using mass spectrometry. The derived rates of change of the isotopomers densities are described accurately by a postulated kinetic model.     

金属样品残余氚测量方法

为定量评价氚在结构材料和老龄贮氢材料内部的滞留量,用化学蚀刻法测定不锈钢内部氚浓度大小、分布情况及同位素交换后老龄贮氚铀床的氚滞留量。结果表明,贮氚13年的不锈钢样品中氚主要存在于样品内表面由表及里的120μm范围内,样品蚀刻深度110.6μm范围内,不锈钢的平均氚滞留量～9.37×10-4mmol/g,贮氚铀粉平均氚滞留量～4.16×10-5mmol/g。该方法对测量金属中微量氚有较高灵敏度,可检测金属中残余氚的滞留量。     

Fuel hydrogen retention of tungsten and the reduction by inert gas glow discharges

Polycrystalline tungsten was exposed to deuterium glow discharge followed by He, Ne or Ar glow discharge. The amount of retained deuterium in the tungsten was measured using residual gas analysis. The amount of desorbed deuterium during the inert gas glow discharge was also measured. The amount of retained deuterium was 2–3 times larger compared with a case of stainless steel. The ratios of desorbed amount of deuterium by He, Ne and Ar glow discharges were 4.6, 3.1 and 2.9%, respectively. These values were one order of magnitude smaller compared with the case of stainless steel. The inert gas glow discharge is not suitable to reduce the fuel hydrogen retention for tungsten walls. However, the wall baking with a temperature higher than 700 K is suitable to reduce the fuel hydrogen retention. It is also shown that the use of deuterium glow discharge is effective to reduce the in-vessel tritium inventory in fusion reactors through the hydrogen isotope exchange.     

Design of a multipurpose laboratory scale apparatus for the investigation of hydrogen isotopes in PbLi and permeation technologies

The knowledge of the tritium transport parameters in lead lithium is fundamental for the design of the HCLL (helium cooled lead lithium) blanket. In fact, the inventory of tritium in fusion reactors blankets and the permeation of tritium into the blanket coolant, with the consequent leaks toward the environment, are strongly depending on its solubility and diffusivity in the lead alloy PbLi. Several experiments, devoted to investigate the function linking the tritium solubilised in lead lithium with the corresponding tritium partial pressure at equilibrium, were carried out in the past, but significant uncertainties still remain.A detailed analysis of the past experimental works is carried out in this paper with the aim to investigate the main problems occurred in the facilities used to measure the tritium solubility in PbLi that caused such a big spread in the achieved results. On the basis of this analysis, a new a multipurpose laboratory scale apparatus has been designed. The apparatus is able to measure the tritium solubility and diffusivity in PbLi in the range of temperature 300–550 °C and it will be operated with hydrogen partial pressure in the range 10²–10⁴ Pa. The facility can work with desorption and absorption technique.Moreover, the apparatus has been designed to allow the testing of H/D concentration sensors in Pb–15.7Li in operative conditions relevant to the HCLL–TBM and the characterisation of hydrogen permeation barrier.     

First wall surface problems for a D ↽ T tokamak reactor

The effects of deuterium, tritium, helium and neutron bombardment on surface degradation of the first wall of a 5000 MW_th D-T reactor have been analyzed. The effects of both sputtering and blistering have been analyzed and the results applied to 316 stainless steel wall operating at temperatures from 300 to 500°C. It has been calculated that the total wall erosion rate is 0.22 mm/year and that 14 MeV neutron sputtering accounts for two thirds of this number. Sputtering from all neutrons results in $\text{≈0.17}\text{mm/year}$ erosion. The calculated erosion rate is 2–3 times that which would be allowable for a 30 year first wall lifetime.     

Helium dilution effect on hydrogen permeation in 316L stainless steel and nickel-base heat-resistant alloys

Effects of inert-gas dilution on hydrogen permeation have been investigated in 316L stainless steel, Inconel 600, Inconel 750, Nimonic 80A and Hastelloy X at 1173 K and 1073 K, by employing a gas-flow system. We used gas mixtures of hydrogen and helium, whose hydrogen concentration ranged from 10^?5 to 10^?1. For the steady-state permeation, the dilution of hydrogen caused no anomalous effects and the permeation rate conformed to Sieverts' law. However, for the transient state, the hydrogen permeation was retarded by the dilution with helium. The retardation effect is discussed in terms of an adsorption model and explained by a decrease in sticking probability at the alloy surface with the dissociative adsorption of hydrogen.     

Effect of Tritium and Dissolved Oxygen on Anodic Polarization of SUS304 Stainless Steel in Sulfuric Acid Solution

Since exotic corrosion of stainless steels in tritiated water can be expected, the anodic polarization of a SUS304 stainless steel sample in approximately 5 wt% sulfuric acid solution was performed at various concentrations of tritium and dissolved oxygen (hereafter DO) in the electrolyte. The inhibitory effect of tritium on the passivation could be observed with DO even at a tritium concentration in the electrolyte of as low as 2.2 kBq cm^?3. This effect became more pronounced as the tritium concentration increased. It was suggested that the inhibitory reaction depending on tritium concentration would compete with the self-passivation depending on the DO concentration (hereafter [DO]), since it was found that there is a threshold [DO] for self-passivation at each tritium concentration.     

Isotope exchange reaction between tritiated water and hydrogen on SiC

SiC has been considered as a primary candidate material for a first wall component in future fusion reactor because it has been claimed that SiC has excellent high-temperature properties, good chemical stability and low activation. However, the behavior of tritium on SiC has not been discussed yet. In this study, tritium trapping capacity on the surface of SiC was experimentally obtained at the temperature range of 25-800 °C in consideration of tritium trapping to the experimental system. The capacity, which was independent of the water vapor pressure in the gas phase and the temperature, was determined as about 10⁶ Bq/cm². The isotope exchange reaction rate between tritiated water in a gas phase and hydrogen on the surface was quantified at the temperature of 25, 500 and 700 °C in consideration of the behavior of tritium trapping at change of experimental condition by the numerical curve fitting method applying the serial reactor model. The reaction rate was observed to be constant as 3.48 × 10⁻⁵ m/s. Additionally tritium release behavior from the surface of SiC in water vapor atmosphere was predicted and compared with that for graphite and stainless steel.     

Tritium Breeding Capability of Heliotron-H Fusion Reactor Blankets

Preparation and operation procedures of chromatographic column for hydrogen isotope separation have been examined. The best separation of isotopic molecular hydrogen was obtained when the stationary phase was activated at 230°C for 16 h and subsequently deactivated with CO₂ at ?7°C. The technique has been applied to analyzing commercially available tritium gases. Protium tritide (HT), DT, and tritiated-methane and -ethane were observed as impurities in all three samples analyzed. It was experimentally confirmed that most of the contaminant protium in the tritium gas came from the inner-surface of the storage vessel.     

Hydrogen transport and solubility in 316L and 1.4914 steels for fusion reactor applications

Equations are given which describe the permeation rate, diffusivity and solubility of hydrogen over the range 250–600°C at pressures up to 10⁵Pa for the 316L stainless and modified 1.4914 martensitic candidate steels proposed for the construction of the Next European Torus (NET). For heat-treated 316L steel, the permeation rates measured agreed well with previous work and did not vary significantly from specimen to specimen or from batch to batch.

Measurements of the permeation rate of hydrogen and deuterium through the modified 1.4914 steel, believed to be the first made, show that the martensitic steel is significantly more permeable than the austenitic steel, by an order of magnitude at 250°C and a factor of five at 600°C. This difference could make it necessary to use permeation barriers on critical components made from the martensitic steel in order to reduce the tritium permeation rate to acceptable levels.     

The long-term behaviors of passivation and hydride layer of commercial grade pure titanium in TRU waste disposal environments

Preservation of the passivity under reducing environmental conditions for extended periods of time and the behavior of hydrogen evolution as the results of the preservation of the passivity of several candidate commercial grade pure titanium related to the small amount of palladium addition, such as Ti–Gr.17 for metallic containers to be buried under deep ground for disposing of transuranic (TRU) waste is investigated. The present investigation has revealed the following corrosion paths for the titanium alloys investigated. The passivity of the alloys is preserved as the result of repeated destruction and recovery of the surface films on the alloys. The long-term corrosion rate under the preserved passivity is of the order of 10⁻⁶–10⁻⁸ my⁻¹ with evolution of hydrogen. The substrate alloys absorb parts of the hydrogen generated to form lath-type hydride phase before forming hydride layers at the final stage.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Interaction of fast H1 + ions with metal surfaces in very high vacuum

The results of experiments on the interaction of fast hydrogen ions (H₁ ⁺) with metals forming weak chemical bonds (nickel, stainless steel) and metals forming strong chemical bonds (tantalum and titanium) with hydrogen are presented. The weighing method was used under very-high vacuum conditions to determine the sputtering coefficient of stainless steel bombarded by 35-keV H ₁ ⁺ ions and the penetration coefficient of H ₁ ⁺ ions entering the stainless steel (=9.10^–3, =0.5 for hydrogen concentrations greatly exceeding 10¹⁹ atoms/cm²). The variation of with the density of the hydrogen introduced and the temperature of the metals was determined by the pressure-variation method. The results indicate that metals of the titanium type are suitable for use in capturing fast hydrogen atoms in magnetic traps.Translated from Atomnaya Énergiya, Vol. 21, No. 5, pp. 339–345, November, 1966.