Kazunari Katayama

Release Behavior of Bred Tritium from Irradiated Li4SiO4

A model to explain tritium release behavior from irradiated Li4SiO4, in the model reported so far by the present authors, it is required to use so small reaction rates for the surface reactions as one several thousandth of the observed values reported in the previous papers to get the good fitting. In this study the mass transfer resistance between grain surface and surface water is newly introduced because it is preferable to use the same reaction rate as that reported previously. The estimated values using the new model give good agreement with the observed tritium release curves and also with the release curves estimated using the model so far.

Isotope Effects of Hydrogen Isotope Absorption and Diffusion in Li0.17Pb0.83 Eutectic Alloy

Abstract Rates and amounts of absorption and desorption of hydrogen and deuterium in a Li0.17Pb0.83 eutectic alloy are determined under the conditions of temperature of 400-700°C and the upstream H2 or D2 partial pressure of 103Pa-105 Pa by using a one-dimensional permeation pot. Because of small interaction between the alloy and dissolved atoms of hydrogen isotopes, the temperature dependence of the Sieverts’ solubility constant for the Li0.17Pb0.83 -H or -D system, i.e., the enthalpy change of absorption or desorption, is small, and the absolute value of D solubility is higher than that of H. The isotope effect of diffusivity between H and D is very small. The generation rate and inventory of tritium (T) in a fusion blanket is estimated under an assumption of one-dimensional Li0.17Pb0.83 blanket system with a constant and uniform neutron flux.

Concentration profiles of tritium penetrated into concrete

Abstract Concentration profiles of tritium in cement paste, mortar and concrete were measured after exposure to tritiated water vapor for a given time. Tritium penetrated a distance of about 5 cm from the exposed surface during an exposure of 6 months. The model of tritium behavior in concrete materials reported by the present authors was developed in this study with the consideration of the effects of sand and aggregate on both the diffusion coefficient of tritiated water vapor and the isotope exchange capacity. Predictive calculations based on the tritium transport model were also carried out in some situations of tritium leakage. The results of the calculations show that a large amount of tritium will be trapped in the concrete walls, and the trapped tritium will be gradually released back to the tritium handling room over the time of months to years even after the decontamination of the room is completed.

Deuterium and Helium Release and Microstructure of Tungsten Deposition Layers Formed by RF Plasma Sputtering

Abstract It is important to evaluate tritium behavior in tungsten deposition layers considering a long-term plasma operation. In this study, tungsten deposition layers were formed by deuterium or helium RF plasma sputtering. The release behavior of deuterium or helium from the layers were observed by a thermal desorption method. When a tungsten deposition layer does not contain oxygen, the retained deuterium is mainly released as D2. When oxygen exists in the layer, the majority of deuterium is released as water vapor. Tungsten deposition layers have an amorphous structure and consist of fine grain with size of 2-3 nm. Numerous bubbles are observed in the layers. A formation of tungsten deposition layer in a fusion reactor may make tritium control more difficult.

Estimation of Tritium Permeation Rate to Cooling Water in Fusion DEMO Condition

Abstract The approximate estimation of tritium permeation rate under the acceptable assumption from a safety point of view is surely useful to progress the design activities for a fusion DEMO reactor. Tritium permeation rates in the blanket and the divertor were estimated by the simplified evaluation model under the recent DEMO conditions in the water-cooled blanket with solid breeder as a first step. Plasma driven permeation rates in tungsten wall were calculated by applying Doyle & Brice model and gas driven permeation rates in F82H were calculated for hydrogen-tritium two-component system. In the representative recent DEMO condition, the following tritium permeation\ rates were obtained, 1.8 g/day in the blanket first wall, 2.3 g/day in the blanket tritium breeding region and 1.6 g/day in the divertor. Total tritium permeation rate into the cooling water was estimated to be 5.7 g/day.

Design Strategy and Recent Design Activity on Japan’s DEMO

Abstract The Joint Special Design Team for Fusion DEMO was organized in 2015 to enhance Japan’s DEMO design activity and coordinate relevant research and development (R&D) toward DEMO. This paper presents the fundamental concept of DEMO and its key components with main arguments on DEMO design strategy. Superconducting magnet technology on toroidal field coils is based on the ITER scheme where a cable-in-conduit Nb3Sn conductor is inserted in the groove of a radial plate. Development of cryogenic steel with higher strength is a major challenge on the magnet. Divertor study has led to a baseline concept based on water-cooled single-null divertor assuming plasma detachment. Regarding breeding blanket, fundamental design study has been continued with focuses on tritium self-sufficiency, pressure tightness in case of in-box LOCA (loss of coolant accident) and material compatibility. An important finding on tritium permeation to the cooling water is also reported, indicating that the permeation to the cooling water is manageable with existing technology.

Hydrogen Retention in a Tungsten Re-Deposition Layer Formed by Hydrogen RF Plasma

Tungsten is a candidate material for plasma facing components for a fusion reactor. Although many studies on hydrogen behavior in tungsten have been carried out, there is insufficient database for a tungsten re-deposition layer. We have made a tungsten re-deposition layer by a sputtering method using a hydrogen and deuterium RF plasma and have investigated hydrogen retention in the layer and the distribution of the layer in the vacuum chamber. The amount of deposited tungsten increased 2.4 times with varying RF power from 100 W to 250 W. It was found from the SEM observation on the cross section that the formed layer has a columnar structure. At high energy (RF power: 250W), a lot of blisters were observed on the surface. The ratio of hydrogen atoms to tungsten atoms (H/W) in the layer was observed to be 0.1 ~ 0.4 with varying RF power. These values of hydrogen retention were much larger than that for absorption into tungsten. Tritium inventory in a D-T fusion reactor may become larger than expected by the formation of tungsten redeposition layer.

Evaluation of Tritium Confinement Performance of Alumina and Zirconium for Tritium Production in a High-Temperature Gas-Cooled Reactor for Fusion Reactors

Abstract Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety and recovery efficiency, tritium confinement is an important issue. It is known that alumina has high resistance for gas permeation. In this study, hydrogen permeation experiments in commercial alumina tubes were conducted and hydrogen permeability, diffusivity and solubility were evaluated. By using obtained data, tritium permeation behavior from an Al2O3-coated Li-compound particle was simulated. Additionally, by using literature data for hydrogen behavior in zirconium, an effect of Zr incorporation into an Al2O3 coating on tritium permeation was discussed. It was indicated that the majority of produced tritium was released through the Al2O3 coating above 500 °C. However, it is expected that total tritium leak is suppressed to below 0.67 % of total tritium produced at 500 °C by incorporating Zr fine particles into the inside of Al2O3 coating, assuming tritium pressure inside particle is kept at the plateau pressure of the Zr hydride generation reaction.

Isotope Effect in Hydrogen Isotope Exchange Reaction on First Wall Materials

Hydrogen isotope exchange reaction is an essential reaction to understand tritium behavior in a fusion reactor because non-negligible amount of tritium is considered to be trapped to the surface of various materials through the isotope exchange reaction. However, a quantitative consideration of the isotope exchange reaction on candidates for the first wall material has not been done yet. In this study we experimentally quantify the hydrogen isotope exchange reaction rates between hydrogen isotope in a gas phase and hydrogen isotope on the surface of isotropic graphite, C/C composite and SiC. The isotope effect is observed in every material used in this work. The reaction rate between H2-gas and D- surface is faster than that between D2-gas and H-surface in every material, though no isotope effect has been observed on the surface of solid breeder materials or metals as stainless steel, aluminum and copper in previous studies. It is also certified that the isotope effect on SiC surface is the largest among three materials studied in this work. The isotope effect in the isotope exchange reaction for various combination of hydrogen isotopes are estimated using Bigeleisen)s equation.

Release Behavior of Tritium from Graphite Material

Abstract The behavior of tritium at removal from graphite material for a fusion reactor is discussed. The mass transfer coefficient representing the isotope exchange reaction between hydrogen isotopes in the gas stream and tritium existing on graphite surfaces and that between water vapor in the gas stream and tritium on the surface are quantified. It was found that the reaction rate between hydrogen isotopes in the gas stream and tritium on the surface is much slower than that between water vapor in the gas stream and tritium on the surface. And, the release behavior of tritium from a graphite particle to the gas phase is calculated with the reaction rates obtained in this study using the solubility and the diffusion coefficient of hydrogen isotopes in graphite that have been presented in the previous report by the authors. A way to remove tritium from a graphite surface applying the isotope exchange reaction between water vapor in the gas stream and tritium on the surface turns out to be effective at the room temperature, although a temperature >1000 K is needed to release tritium from the bulk of a 10-μm graphite particle.