Ratio of Permeabilities of Tritium to Protium through Palladium-Alloy Membrane

Experiments were performed to study the ratio of permeabilities of protium to tritium through Pd-Ag-Au alloy at the conditions of 350°C≦T≦550°C and p<1,000 mmHg.

The observed ratio was 2.12±0.03 at 350°C, and it was almost independent of temperature. It is remarkable that the obtained ratio of permeabilities is considerably larger than the square root of the atomic weight ratio?3.

The dependence of separation factor for the hydrogen isotopes on pressure and “cut” is also discussed.     

Detritiation of Glovebox Atmosphere by Using Compact Tritium Removal Equipment

A compact tritium removal equipment (TRE), assembled in a console with casters, has been developed for detritiation of air in a glovebox used for handling of several curies of tritium. The TRE was designed to remove gaseous tritium in the form of T₂, HT and CH₃T through oxidation with precious metal/alumina catalysts followed by adsorption on zeolite pellets.

From the detritiation experiments with hydrogen tritide (HT, 2–20 mCi), the TRE was confirmed to have sufficient performance for the practical use. The tritium concentration in the test gas (total volume –32l; 1%H₂, 5%O₂, 94%N₂) decreased from 0.64 to 6.4 ×10-⁷ Ci.m³ within 155 min when the TRE was operated under the recirculation mode with the flow rate of 200 l-h¹ at the catalyst temperature of 200°C. In addition, the HT-to-HTO fractional conversion was determined at various catalyst temperatures (25–200°C) and flow rates (100–360 lh-¹).     

Tritium diffusion through oxide surface films on niobium

Diffusion of tritium in niobium has been investigated using recoil implantation with particular emphasis on the role of surface oxide films. In conformity with previously reported data, bulk diffusion in niobium is found to be extremely rapid and surface oxide films significantly retard the release rate of tritium. The observed tritium release data indicate diffusion through cracks or defects in the oxide film below 500°C and transport through an intact film at temperatures higher than 500°C. The release rates are described in terms of a mathematical model with the necessary model parameters determined from experimental data. The oxide films could be removed with a reducing hydrogen atmosphere, thereby resulting in classical bulk diffusion controlled release above 700°C. Below 700°C the oxide layers were stable in hydrogen, presumably because of a moderately high oxygen potential resulting from minute amounts of moisture in the sample chamber.     

Kinetic estimation of hydrogen isotope exchange reaction between tritiated-water (HTO) vapor and each amino acid in a heterogeneous system

The hydrogen isotope exchange reaction between HTO vapor and each amino acid has been observed in order to establish a method of estimation of the internal exposure of organically bound tritium in a heterogeneous system at 25～70°C. Rate constants (k) for the amino acids have been obtained by applying the A″-McKay plot method. Using these k values, Arrhenius plots for both the COOH and NH₂ groups are drawn, and linearity was obtained over the range of 25～70°C. Comparing the rate constants, the following four statements can be made regarding the T-for-H exchange reaction. (1) The reactivity of the functional groups in amino acids increases with increasing temperature. (2) Applying Taft's equation, the ratio of polar effect to steric effect is 2:8 in the COOH group and 9:1 in the NH₂ group at 25°C. (3) The A″-McKay plot method is useful for studying the reactivity of materials, not only with one (or the same kind of) functional group(s) but also with two different kinds of functional groups. (4) The method used in this work may be useful to investigate the behavior of organically bound tritium, quantitatively.     

Permeation Analysis of Tritium Through the Vessel Wall of a Titanium Hydride Storage Container

Abstract

A preliminary design for a stainless steel vessel for the long-term storage of hydrogen isotopes has been proposed. The immobilised hydrogen, as a titanium hydride, could be stored in a stainless steel vessel for this application. The vessel, as a primary package, is designed to form titanium hydride and to contain the hydrogen isotopes and helium-3 produced from the decay of tritium. In order to predict the possibility of contamination and the deterioration of the mechanical properties, a numerical diffusion analysis calculation of the hydrogen isotopes and helium inside the stainless steel vessel was carried out. Numerical results showed that a negligible amount of tritium would be released by permeation through a 0.7 cm thick vessel wall at normal conditions over the entire period of the storage. When the vessel is heated up to a temperature of 600^°C for the routine conditions of activation or exothermic hydriding, tritium loss or contamination would be of little concern. However, if the vessel were exposed to fire conditions with a temperature of 800^°C, permeation of hydrogen through the vessel wall would result in a serious increase in the amount of tritium escaping, in a very short time.     

Tritium Depth Profiling Study on Titanium Tritide Target for Generating 14-MeV Neutrons

In-situ measurements of the depth profiles of tritium in a titanium tritide target for generating 14-MeV neutrons have been made with the method of the ion beam analysis using the T(d, α)n nuclear reaction. The initial distribution of tritium in the unirradiated target has been observed to be nearly uniform over the depths. After the irradiation of 390-keV D₃ ⁺ ions at a temperature of about 10°C a dip has been found in the depth profile around the depth of the projected range of the ions. By the successive isochronal anneal-ings at temperatures below 130°C the tritium has been uniformly redistributed. The behavior of tritium in the target and the effectiveness of the depth profiling for evaluating the energy spectrum and the yield of source neutrons are discussed.     

Decomposition pressures in the (α + β) fields of the Li-LiH,Li-LiD,and Li-LiT systems

Decomposition (“plateau”) pressures of H₂, D₂, and T₂ over Li-LiH, Li-LiD, and Li-LiT alloys were determined between 600 and 850°C, for pressures ranging from 3 to 460 Torr, and for alloy compositions falling within the (α + β) miscibility gaps. The measurements were carried out separately for each hydrogen isotope using the same lithium sample and experimental procedures. For each system the ln P vs. 1/T data form a pair of linear segments, the intersection of which represents the monotectic temperature (694°C for Li-LiH, 690°C for Li-LiD, and 688°C for Li-LiT). For a given temperature plateau pressures were in the order P_T2 >P_D2 >P_H2. The $\text{D}\text{H}$ and $\text{T}\text{H}$ isotope effects in pressures varied from 1.48 and 1.74 at 600°C to 1.34 and 1.45 at 850°C. The results were used to calculate the standard free energies of formation of solid LiH, LiD, and LiT. The tritium gas used in this study had significant amounts of hydrogen and deuterium. A method for correcting the plateau pressures of this mixture to those of pure tritium is presented.     

Hydrogen Extraction Characteristics of Proton-conducting Ceramics under a Wet Air Atmosphere for a Tritium Stack Monitor

For the purpose of designing a tritium monitoring system combined with proton-conducting ceramics as a membrane separator, the hydrogen pump characteristics of CaZr_0.9In_0.1O_3–α proton-conducting ceramics were evaluated. In the experiments, argon gas containing 20.7% oxygen and 1.2% water vapor was fed to the anode at a rate of 47–137 ml/min at 600–800°C and an applied voltage until 3.5 V. The resulting hydrogen evolution rate reached maximally 0.67ml/min and the hydrogen recovery rate was 60%. However, the proton transport number decreased to 0.52 because the electron-hole current increased along with protonic current according to the defect equilibrium reaction occurring under a wet atmosphere containing oxygen. During operation, the hydrogen evolution rate fluctuates over time by at least 0.1 ml/min, which is approximately 20% of the hydrogen evolution rate. Additionally, the hydrogen evolution rate increased with an increase in the partial pressure of water vapor at the anode. It is important to design the tritium monitoring system taking into consideration the fluctuation in hydrogen evolution rate.     

Evaluation of the response time of H-concentration probes for tritium sensors in lead–lithium eutectic alloy

Dynamic tritium concentration measurement in lead–lithium eutectic is of major interest for a reliable tritium testing program in ITER TBM and for an experimental proof of tritium self-sufficiency in liquid metal breeding systems. Potentiometric hydrogen sensors using different solid-state electrolytes for molten lead–lithium eutectic have been reported and tested by the Electrochemical Methods Lab at Institut Quimic de Sarria (IQS).In the present work the following ceramic elements have been synthesized and characterized by X-ray diffraction (XRD) in order to be tested as a Proton Exchange Membranes (PEM) H-probes: BaCeO₃, BaCe_0.6Zr_0.3Y_0.1O_3−δ and Sr(Ce_0.9–Zr_0.1)_0.95Yb_0.05O_3−δ. Potentiometric measurements of the synthesized ceramic elements have been performed shifting from a fixed hydrogen partial pressure at the working electrode to high purity argon. In this experimental campaign a fixed and known hydrogen pressure has been used in the reference electrode. The goal of these experiments is to evaluate the sensor response time when the hydrogen concentration in the environment is rapidly changed. All experiments have been done at 500 °C and 600 °C. The sensor constructed using the proton conductor element BaCe_0.6Zr_0.3Y_0.1O_3−δ exhibited stable output potential and its value was close to the theoretical value calculated with the Nernst equation. In contrast, the sensors constructed using the proton conductor elements BaCeO₃ and Sr(Ce_0.9–Zr_0.1)_0.95Yb_0.05O_3−δ showed higher deviations between experimental and theoretical data, and long response times.     

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.     

Separation Factor of Hydrogen and Deuterium by Suspension Electrolysis Using Mercury Cathode

The electrolytic separation factor between hydrogen and deuterium was examined using mercury or else platinum cathode immersed in IF NH₄Cl in 10v/oD₂O water containing cobalt sulfide powder in suspension. Several other kinds of powdered materials in suspension were also studied. In the case of mercury pool electrode, the materials added in suspension were effective in enhancing the hydrogen/deuterium separation factor, but powder suspension was ineffective on platinum plate electrode. The powdered material added in suspension served as catalyst on the hydrogen evolution reaction at the mercury cathode. The influence of the applied potential on the separation factor was studied over the temperature range of 15°~80°C. The results provided an indication of the rate-determining steps governing the electrolytic hydrogen evolution.

The experimental values obtained for the separation factor and activation energy gave an insight into the mechanism of the rate-determining step of the hydrogen evolution at the working electrode.     

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

The tritium concentration in the hydrogen product in Japan's future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO₃ decomposer and a H₂SO₄vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10^?3 Bq/cm³ at STP (38 Bq/g-H₂) to 0.18 Bq/cm³ at STP (2,000 Bq/g-H₂). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10^?5 Bq/cm³ at STP (0.36 Bq/g-H₂) to 5.6 × 10^?3 Bq/cm³ at STP (63 Bq/g-H₂), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10^?2 and 9.6 × 10^?3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.     

Development of a plasma driven permeation experiment for TPE

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1], [2]. To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.     

Preliminary results from a detritiation facility dedicated to soft housekeeping waste

Nuclear waste management has to be taken into account for fusion machine in tritium experimentations. Soft housekeeping waste is produced during both operating and dismantling phases and is contaminated by tritium under reduced (HT) and oxidized (HTO) forms. At CEA Cadarache, a lab-scaled facility has been built for soft housekeeping detritiation. The tritiated gas exhausted from the process described above is foreseen to be treated by a tubular Pd–Ag membrane reactor, for gaseous tritium recovery. Since this membrane reactor uses hydrogen as swamping gas the compatibility toward explosive hazard has to be taken into account. Then, this work presents a double objective. A first study is presented in order to identify the best conditions for the declassification of soft housekeeping waste, without tritium recovery. Experiments carried out at 120 °C are not efficient enough and do not allow one to choose the most efficient carrier gas. Some other tests are being currently performed at higher temperatures (150 °C). Moreover, due to safety issues, the use of air has to be avoided during membrane reactor implementation phase. Preliminary results obtained with hydrogen hazard-free carrier gases are also presented.     

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.     

Tritium release experiments on Li4SiO4 pebbles from TRINPC-I experiments: Effects of water adsorption and hydrogen gas

Out-of-pile tritium release experiments under different water uptake contents and purge gas chemistry were performed on Li₄SiO₄. Water measurement was performed on samples under different experimental procedures. It was found that water was adsorbed on the sample during its transferring and storage process. A strong dependence of tritium release behavior on water uptake was determined. By doping H₂ in the sweep gas, the formation of water in orthosilicate was observed in addition to the isotope exchange reaction with H₂ gas. Thermal desorption peaks of the water formation reaction and H₂ isotope exchange reaction appeared at 668 °C and 463 °C, respectively, at ramping rate of 5 °C/min.     

Studies on the Characteristic of Zirconium–Tritium Reaction

The p–c–T curves of tritium absorption and desorption of zirconium 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 range from 450 to 500°C and two plateaus at temperature above 600°C. The thermodynamic parameters of the different phases were determined according to the van’t Hoff equation. The hysteresis effect was observed in reversible process of tritium absorption and desorption of zirconium on our experimental condition. The tritium absorption behavior by zirconium in the temperature range from 450 to 620°C and desorption behavior of zirconium in the temperature range from 775 to 875°C have been investigated. 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 (−16.8 ± 0.8) kJ·mol⁻¹ and (57.7 ± 1.6) kJ·mol⁻¹, respectively.     

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.     

Hydrogen Isotope Separation by Plasma-Chemical Method

Deuterium transfer (exchange) reaction as shown in HDO+H₂=H₂O+HD was studied as a case similar to the tritium transfer reaction as shown in DTO+D₂=D₂O+DT, the first step in tritium isotope separation of the tritiated heavy water DTO. The transfer reaction was plasma-chemically catalyzed by allowing a gas mixture such as H₂O/D₂, D₂O/H₂, H₂O/HDO/H₂ or H₂O/D₂O/HDO/H₂ to flow through an atmospheric pressure discharge zone formed in a reaction chamber, the inner temperature of which was maintained just above 100°C. The plasma-chemical reactions diagnosed by infrared and emission spectroscopy revealed that the mixture underwent instantaneous deuterium transfer reactions as it passed the zone. The effectiveness of the method was demonstrated by high deuterium transfer rate, high separation factor of the transfer and a possibility of miniaturizing the separation facility.