Futaba Ono

Storage of Hydrogen Isotopes in Uranium Alloys

Hydrogen absorption-desorption properties of some uranium-rich alloys were studied to develop an advanced material for temporary tritium storage. The characteristics of U-33.3 at%Zr were analogous to those of U-69.7at%Zr, while the durability to powdering was reduced according to a decrease of zirconium content. The U-5.7at%Al exhibited a similar behavior to that of pure uranium including powdering to much extent, though a slight decrease in decomposition pressure was observed. The U-25at%Si showed very small capacity of hydrogen in spite of the existence of uranium phase in this alloy.

Sorption and Desorption of Tritiated Water on Paints

In this paper the sorption and desorption of tritiated water on paints, such as acryl, silicone, epoxy and fluorine types, has been studied. In dynamic adsorption and desorption experiments, tritium concentration in the chamber, in which painted metal was installed, was measured during HTO vapor charging or purging by N{sub 2} + H{sub 2}O. Slow desorption rate for acryl type was observed. Residual tritium not released by sweeping with N{sub 2} + H{sub 2}O was recovered by heating up to 800 C. Residual tritium fraction for epoxy type was found to be highest among tested. Paint membrane was used in permeation experiments of HTO vapor and liquid. Diffusivity, solubility and permeability were evaluated. Diffusivity for fluorine was largest and that for epoxy smallest.

Removal of tritiated water vapor by adsorption on molecular sieves - effects of co-existing H/sub 2/O

Effect of co-existing H/sub 2/O (isotopic swamping effect) in the adsorption of tritiated water vapor on molecular sieves was experimentally studied by two means. One was by breakthrough experiment. It was found that H/sub 2/O-HTO two components Langmuir type adsorption isotherm is satisfactory to explain the HTO breakthrough curve. The other was by using a divisible column, where behavior of HTO at much lower concentration than at the inlet and the axial distribution of tritiated water on the adsorbent were studied. Concentration of tritiated water vapor at the column outlet was found to decrease with the increase of H/sub 2/O concentration of the input. However the improvement of decontamination factor was little.

Tritium sorption by cement and subsequent release

Abstract In a fusion reactor or tritium-handling facilities, contamination of concrete by tritium and subsequent release from it to the reator or experimental room is a matter of problem for safe control of tritium and management of operational environment. In order to evaluate this tritium behavior, interaction of tritiated water with concrete or cement should be clarified. In the present study, HTO sorption and subsequent release from cement were experimentally studied. 1. (1) Sorption experiments were conducted using columns packed with cement particles of different sizes. From the analysis of the breakthrough curve, tritium diffusivity in macropores and microparticles were evaluated. 2. (2) From the short-term tritium release experiments, effective desorption rate constants were evaluated and the effects of temperature and moisture were studied. 3. (3) In the long-term tritium release experiments to 6000 h, the tritium release mechanism was found to be composed of three kinds of water: initially from capillary water, and in the second stage from gel water and from the water in the cement crystal. 4. (4) Tritium release behavior by heat treatment to 800°C was studied. A high temperature above 600°C was required for the tritium trapped in the crystal water to be released.

Simulation of soaking effect in air detritiation system

Overall efficiency of air detritiation operations in the fusion reactor containment is affected by several factors, which include conversion of gaseous tritium to tritiated water vapor, sorption by materials such as wall and floor, and consequent desorption from these materials back into the reactor containment. Simulation study of air detritiation systems has been performed by bench-scale experiments and computer modeling. In the case that gaseous tritium is released into the containment, high-temperature metal surface has a significant effect on generating tritiated water vapor and thus generated tritiated water vapor mainly causes the delay of the required operation time. Tritium concentration reduction curves were found to depend on the characteristics of the materials installed in the containment in relation to conversion of HT to HTO at the surface, adsorption/desorption and solution/diffusion. These dependences were also shown by the model calculation.

Study on surface defect structures of ZrO2 and some doped ZrO2 by means of work function measurement

Abstract The work function change of the ZrO 2 + 2%Y 2 O 3 sintered pellet, caused by a change of the composition of the sweep gas, was measured using a high temperature Kelvin probe. The Pt reference electrode was calibrated by using ZrO 2 + 2%Y 2 O 3 as a standard material. Work function changes of undoped ZrO 2 and Nb-doped ZrO 2 (2%Nb 2 O 5 ) were measured as a function of equilibrium oxygen partial pressure, P O2 . The thus obtained exponents of P O2 , 1/ n , were 1/6.2 and 1/33.6 for ZrO 2 and ZrO 2 + 2%Nb 2 O 5 , respectively. These exponent values were discussed in terms of defect chemistry of the surface layer.

Preparation, analysis and irradiation of hydrided U–Th–Zr alloy samples for a new fuel

Abstract To get the properties of U–Th–Zr–H alloys, which are needed to utilize them for nuclear fuels, changes in the dimensions and weights of the alloys on hydrogenation and in microstructure and hardness on neutron irradiation to 7.4×1023 n/m2 were examined. The hydrogenated alloys show high apparent densities and high durability for the irradiation, which promotes the utilization of the alloys for a new type of nuclear fuel.

Study of the retention of hydrogen isotopes implanted in Mo

A thin foil of Mo was irradiated by ions of hydrogen isotopes including tritium using a tritium beam test apparatus at the University of Tokyo. Thermal desorption spectroscopy (TDS) technique was employed to evaluate the total amount of retained hydrogen isotopes in Mo. According to the results, deuterium retention increased monotonously as the irradiation fluence was increased. It was observed that thermal desorption of both hydrogen and deuterium occurred predominantly at low temperatures, below 400 K, and that the increase of desorption could be attributed to higher temperature peaks which became appreciable as the fluence was increased; but, the amount of hydrogen or deuterium desorption from these high temperature peaks were found to be very small. The tritium retention in Mo, on the other hand, was nearly the same as that found for Ni.

Hydrogen absorption-desorption properties of UCo

UCo absorbs hydrogen to form UCoH2.7 powder below 423 K at 105 Pa. Although it dehydrogenates reversibly below 423 K, two-phase mixtures of UH3 and UCo2 appear above 473 K, which can be explained on the basis of the free energy differences. The pyrophoricity of the hydrogen storage system based on UCo is expected to be lower than that based on pure uranium because the two-phase mixtures recycle to UCo after complete desorption above 673 K and uranium metal rarely appears in the hydrogen absorption-desorption cycle. However, the decomposition pressure at low temperatures needs to be reduced in order to use this compound for tritium storage.

Hydrogen desorption properties of hydrogenated UThZr alloys

Abstract Hydrogen desorption properties of hydrogenated UThZr alloys of varied compositions were investigated using a hydrogen absorption-desorption experimental system, TG-DTA and DSC analyzers. Isothermal desorption at 900°C of elemental ratio U:Th:Zr:H = 1:1:4:9.5 exhibited that there were two distinct plateau regions identified as ZrH 1.4 -ZrH and ThZr 2 H 7− x − ThZr 2 systems. TG-DTA and DSC measurements under the temperature range from room temperature to 1000°C have shown that there were three endothermic peaks identified as dehydrogenation reactions of ZrH 2 − x − ZrH and ThZr 2 H 7−x . The DTA curve identified the first peak area as the ZrH 1.4 −ZrH system, while the DSC curves identified that the second peak is the decomposition of ZrH and the third peak is the decomposition of ThZr 2 H 7 − x . It was also shown that both ZrH 2 − x and ThZr 2 H 7 − x are more stable in the alloy than the pure ones. Measured enthalpy changes during decomposition of the hydrogenated UThZr alloy are similar to the theoretical calculation. Oxidation during measurement of the U:Th:Zr:H = 2:1:6:13.1 resulted in a different measured enthalpy change and calculation. Isothermal decomposition of the U:Th:Zr:H = 1:1:4:9.5 without any disintegration indicates stability of the alloy against powdering on hydriding-dehydriding cycles. Stability of the samples at high temperature similar to that of UZrH 1.6 for TRIGA fuel can be maintained after the decomposition.