Teppei Otsuka

Observation of Hydrogen Distribution Around Non-Metallic Inclusions in Steels with Tritium Microautoradiography

Hydrogen distributions around non-metallic inclusions in steels are successfully characterized with high-resolution tritium autoradiography. The autoradiographs show that hydrogen accumulation characteristics around the inclusions depend on types of the inclusions. In the case of MnS, hydrogen was inhomogeneously distributed in the ferrite matrix surrounding the MnS inclusion, probably because hydrogen is trapped in defects formed around MnS. The inhomogeneous distribution of hydrogen may be originated from the asymmetric stress field produced by a contraction of the MnS phase in the heat treatment, i.e. the inhomogeneous volumetric change of MnS owing to its larger thermal expansion than that of the ferrite phase. In the case of Al2O3, hydrogen was intensely localized at boundary layers of the ferrite matrix surrounding the Al2O3 inclusion. This could be attributed to hydrogen trapping at defects introduced by a residual stress in the boundary layers of the ferrite matrix due to larger contraction of the ferrite phase than that of the Al2O3 phase on cooling. Similarly hydrogen was accumulated in the surrounding ferrite matrix but more widely distributed around Cr carbide probably because difference in the thermal expansion between the Cr carbide and ferrite phases is less than that between the Al2O3 and ferrite phases.

Visualization of hydrogen depth profile by means of tritium imaging plate technique: determination of hydrogen diffusion coefficient in pure tungsten

The tritium imaging plate (TIP) technique has been applied to visualize penetration profiles of hydrogen (tritium) loaded in pure tungsten (W) by a dc glow discharge at a temperature ranging from 473 to 673 K. The penetration profile consists of two components, i.e. a highly localized one in the near-surface region (sub-mm in depth), and another, deep penetrating one (several mm in depth). An apparent hydrogen diffusion coefficient is determined from the latter to be , which agrees well with the extrapolation of Frauenfelders data obtained at elevated temperatures. The near-surface localized one is attributed to hydrogen trapping with a trapping energy of 0.84±0.04 eV.

Application of Tritium Tracer Technique to Determination of Hydrogen Diffusion Coefficients and Permeation Rate near Room Temperature for Tungsten

Abstract Applying a tritium tracer technique, we have investigated hydrogen plasma driven permeation (PDP) through tungsten (W) near room temperature. The technique was confirmed to give reliable data on diffusion and permeation coefficients of pure W for gas driven permeation (GDP), and then it was applied to observe PDP in W near room temperature. It was found that PDP in earlier phase was controlled by diffusion giving reliable diffusion coefficients. Taking literature data at higher temperatures and present ones near room temperature determined from PDP into account, we have proposed new diffusion coefficients DUpper limit = (3.8±0.4)x10-7 exp ((-39.8±1.5) (kJ/mol)/RT), m2s-1. (1) The activation energy for permeation determined by PDP was similar to that by GDP. The extrapolation of the present data to higher temperature agreed well with Frauenfelder’s data, suggesting the activation energy of around 65 kJ/mol for permeation is quite reasonable. However prolonged measurements resulted in significant reduction of PDP. The cause of the reduction was attributed to the increase of reemission owing to surface cleaning and/or roughening by incidence of energetic hydrogen.

Behaviour of tritium in plasma-sprayed tungsten coating on steel exposed to tritium plasma

In order to understand the role of a plasma-sprayed tungsten (W) coating on tritium (T) permeation in a W-coated ferritic/martensitic steel (F82H), we have examined depth profiles of T in the coating and the substrate using the tritium imaging plate technique after T loading by a dc glow-discharged plasma at 453 and 573 K for 2 h. Tritium loaded by plasma exposure was distributed uniformly in the whole coating, while T penetrated to the substrate by diffusion. The former is caused by T diffusion through open pores and/or along grain boundaries followed by adsorption on grain surfaces and dissolution in the grains. The main role of the W coating on T permeation is to reduce the incoming flux at the coating/substrate interface owing to pore diffusion in the coating and the effective area for T dissolution in the substrate.

HYDROGEN RELEASE FROM FERRITIC/MARTENSITIC STAINLESS STEEL

Abstract Hydrogen release behaviors from the 8Cr2W stainless steel (RAF/M) around RT are examined by using tritium tracer techniques, and trapping effects of bulk and surface are discussed. In the overall release, three different release stages are clearly distinguished giving three different diffusion coefficients and release amounts which indicate the existence of different kinds of trapping. In addition, the appreciable amount of hydrogen (tritium) is trapped on the surface and/or surface oxides of RAF/M, but they are hardly released and show no influence on the overall hydrogen release behavior. At very low hydrogen concentration, almost all hydrogen atoms are trapped at the deepest trapping site, probably M23C6, and the sites are easily saturated. With increasing the hydrogen concentration, the shallower trapping sites are occupied. Remaining hydrogen atoms seem to be in normal (interstitial) sites, whose amount increases with the square root of the hydrogen loading pressure, but they are still influenced by trapping with lattice imperfections and/or grain boundaries.

Diffusion of helium in water-saturated, compacted sodium montmorillonite

1Division of Quantum Energy Engineering, Graduate School of Engineering, Hokkaido University, Kita-13, Nisi-8, Kita-ku, Sapporo 060-8628 2Advanced Energy Engineering Science, Interdisciplinary, Graduate School of Engineering, Kyushu University, Hakozaki, Higashi-ku, Fukuoka 812-8581 3Nuclear & Environmental Technology Development, Research & Development Division, JGC Corp., 2205, Narita-cho, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1313

Change of chemical states of niobium in the oxide layer of zirconium–niobium alloys with oxide growth

The change of chemical states of niobium with oxide growth was examined in the oxide layers of Zr–2.5Nb around the first kinetic transition by the conversion electron yield – X-ray absorption near-edge structure measurements. The detailed depth profiles of niobium chemical states were obtained in both the pre- and the post-transition oxide layers of Zr–2.5Nb formed in water at 663 K for 40–280 d. The depth profiling revealed that the inner oxide layer remained protective to oxidizing species even though in the post-transition region and this excellent stability of barrierness would be attributed the suppression of hydrogen pickup.

Characterization of surface morphology and retention in tungsten materials exposed to high fluxes of deuterium ions in the tritium plasma experiment

Under appropriate conditions, exposing tungsten to a high flux D plasma creates near-surface blisters and other changes in surface morphology. We have characterized the sizes of blisters formed at different temperatures (147 °C≤Tsurface≤704 °C) and performed a surface analysis to elucidate factors that influence blister formation. Tungsten targets that were exposed to low energy (70 eV) D ions at a flux of 1.1×1022 m−2 s−1 in the tritium plasma experiment (TPE) were considered. We used AES to analyze the surface for evidence of implanted impurities. Blister diameters and heights were quantified using SEM imagery and vertical scanning interferometry. Given the likelihood of D precipitation in blisters, we expect that the data obtained here could be incorporated into a computational model to better simulate the diffusion and desorption of D in W. With this in mind, we present an analysis of thermal desorption profiles showing the release of D from the surface.

Overview of Recent Tritium Experiments in TPE

Abstract Tritium retention in plasma-facing components influences the design, operation, and lifetime of fusion devices such as ITER. Most of the retention studies were carried out with the use of either hydrogen or deuterium. Tritium Plasma Experiment is a unique linear plasma device that can handle radioactive fusion fuel of tritium, toxic material of beryllium, and neutron-irradiated material. A tritium depth profiling method up to mm range was developed using a tritium imaging plate and a diamond wire saw. A series of tritium experiments (T2/D2 ratio: 0.2 and 0.5 %) was performed to investigate tritium depth profiling in bulk tungsten, and the results shows that tritium is migrated into bulk tungsten up to mm range.

Behavior of Tritium near Surface Region of Metals Exposed to Tritium Plasma

Abstract In order to understand behavior of tritium (T) on surface and in bulk of metals exposed to T plasma, both surface activities and depth profiles of T were periodically observed by a tritium imaging plate technique during storage in air at room temperature (RT) for over 1 year. In the T depth profiles, T localized within a depth of sub mm from the surface was clearly distinguished from T in the bulk. The former was attributed to strong trapping by some defects produced by the plasma exposure and remained quite longer during the storage, while the latter was released from the surfaces by diffusion. T surface activity measured on the plasma-exposed surface changed in a complicated way with time due to removal of T by isotopic replacement with H in ubiquitous H2O and T supply from the bulk in the course of the diffusional release.