Fluence dependence of deuterium retention in oxidized SS-316
Journal of Nuclear Materials, 417 (2011) 1154-1157
Y. Oyaa, S. Suzukia, M. Matsuyamab, T. Hayashic, T. Yamanishic,
Y. Asakurad, K. Okunoa
aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
bITER Organization, St.Paul-lez-Durance, France
cJapan Atomic Energy Agency, Ibaraki, Japan
d National Institute for Fusion Science, Toki, Japan
Abstract
The ion fluence dependence of deuterium retention in SS-316 during oxidation at a temperature of 673 K
was studied to evaluate the dynamics of deuterium retention in the oxide layer of SS-316. The correlation
between the chemical state of stainless steel and deuterium retention was evaluated using XPS and TDS. Journal of Nuclear Materials, 417 (2011) 576-580 J. Wua, Z. Yanga, Q. Lia, C.-Y. Xiea,
G.-N. Luoa, M. Matsuyamaa
a Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei 230031, China Abstract Tritium exposed samples, including doped graphite, SiC coated graphite, tungsten coatings fabricated by
vacuum plasma spraying, and polycrystalline tungsten, were examined by b-ray-induced X-ray spectrometry
in an argon atmosphere. The changes in the X-ray spectra with time were followed for a maximum
time of 100 h. The results indicated that the SiC coated graphite absorbed the most tritium and the
polycrystalline tungsten the least. Preliminary computer simulations were performed to analyze the
depth profiles of tritium in these materials. Materials Transactions, Vol. 52, No.3 (2011) 526-530 T. Nozaki , H. Homma, Y. Hatano
Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
Abstract Pressure-composition isotherms of TiC1-x -H system were measured
at 773,873 and 973 K over a pressure range from 10-2
to 105 Pa. The obtained isotherms could be divided in three concentration regions.
In the low concentration region in which[H]/[TiC] ≦ 1.4 × 10-3,
Sieverts' law held, and the solubility of hydrogen clearly increased with decreasing temperature.
The enthalpy of solutions was 70 ± 5 kJ mol-1. In the region in which
1.4 × 10 -3 < [H]/[TiC] ≦ 8.0 × 10-3, the dependence
of hydrogen concentration, CH, on pressure and temperature weakened
as pressure lincreased. Finally,CH was independent of temperature when
it reached 8.0 × 10-3. This CH was comparable with the concentration
of carbon vacancies in TiC1-x. In the high concentration region in which [H]/[TiC]
≧ 8.0 × 10-3, Sieverts' law held agaln, but the solublilty of
hydrogen was almost independent of temperature. These observations independent of temperature.
These observations indeicated that hydrogen dissolved in carbon vacancies in the low and
middle concentration regions with the above-mentioned value of enthalpy of solution,
and also occupied much less stable sites in the high concentration region. Journal of Nuclear Materials, 417 (2011) 900-903 M.Matsuyama1, Z.Chen2, K.Nisimura3, S.Akamaru1, Y.Torikai1, Y.Hatano1, N.Ashikawa3, Y.Oya4, K.Okuno4, T.Hino5 1Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama, 930-8355,Japan
2The Southwestern Institute of Physics, Chengdu 610041, Sichuan, China
3National Institute for Science, Toki 509-5292, Japan
4Radiochemistry Research Laboratory, Shizuoka University, Shizuoka 422-8529, Japan
5Laboratory of Plasma Physics and Engineering,Hokkaido University, Sapporo 060-8628, Japan
Abstract
Tritium amount retained in surface layers and release behavior from surface layers were examined using
SS316L samples exposed to plasmas in the Large Helical Device and commercial Cu-Be alloy plate.
BIXS analysis and observation by SEM indicate that carbon and titanium deposited on the plasma-facing
surface of the SS316L samples. Larger amount of tritium was trapped in the plasma-faclng surface in
comparison with the polished surface. Hligher enrichment of tritium in surface layers was similarly found in
the pollShed surihce of both samples.The amount of surface trltlum in both samples was alrnost same,
comparison with polished surface. Higher enrichment of tritium in surface layers was similarly found in
the polished surface of both samples. The amount of surface tritium in both samples was almost same,
while the bulk concentration of tritium in Cu-Be was much lower than that in SS316L. Tritium release from
the SS316L and Cu-Be samples into water was examined by immersion experiments. Tritium elution was
observed for both samples, but changes in the residual tritium amount in surface layers were different from
each other.
Keywords: Tritium, trapping, Stalnless steel, Cu-Bealloy, BIXS International Journal of Hydrogen Energy 36 (2011) 12333-12337
M. Haraa, K. Yudoub, E. Kinoshitab, K. Okazakib, K. Ichinoseb,
K. Watanabea, M. Matsuyamaa
Abstract To study the alloying effects on ZrMn2H system, thermodynamic properties of Zr(Mn1xCo
x)2 hydride were measured by volumetric method. ZrMn2 gave a single plateau region in the pressurecomposition isotherm. On the other hand, double plateaus were clearly observed in Zr(Mn0.7Co0.3)
2 and Zr(Mn0.6Co0.4)2H systems.
The appearance of the double plateau characteristics would be explained in view of the hydrogen binding in the tetrahedral occupation sites in Zr(Mn1xCox)2. Since the hydrogen binding in the tetrahedral 2ZrMnCo site would be less stable than that in the 2Zr2Mn site,
the equilibrium pressure increases with increasing cobalt content. The appearance of the first plateau was ascribed the increase in the bonding of MnH in 2Zr2Mn site adjoining the 2ZrMnCo site.
Keywords: Alloying effect, ZrMn2, Double plateau, Hydrogen occupation site
M. Inoue1Journal of Nuclear Materials, 417 (2011) 1336-1340
P. Calderoniaa, J. Sharpea, M. Shimadaa, B. Dennya,
B. Pawelkoa, S. Schuetza, G. Longhursta, Y. Hatanob,
M. Harab, Y. Oyac, T. Otsukad, K. Katayamad, S. Konishie,
K. Noborioe, Y. Yamamotoe
Abstract The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.
Fusion Engineering and Design, 86 (2011) 1776-1779
Y.Oyaa, M.Kobayashia, R.Kurataa,
N. Yoshidab, N. Ashikawac, A. Sagarac, M.Harad,
Y.Hatanod, K.Okunoa
Abstract
Behaviors of hydrogen isotope retention and damages in tungsten and SS-316 with simultaneous
C+-D2+implantation were compared to those with only D2+ implantation using X-ray photoelectron spectroscopy (XPS), Thermal desorption spectroscopy (TDS), glow discharge-optical emission spectroscopy (GD-OES) and Transmission electron microscopy (TEM).
Keywords: Tungsten, SS-316, Simultaneous implantation, Fuel retention
Fusion Engineering and Design, 86 (2011) 2152-2155
T. Yamanishia, T. Hayashia, Y. Iwaia, K. Isobea, M. Harab,
T. Sugiyamac, K. Okunod
Abstract
The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. It was found that the permeation rate of tritium was almost the same as that with no water vapor, and the major chemical form of tritium was HTO. It was also found that the permeation rate was decreased to ~1/10 of the above value by plated with gold, and the major chemical form was HT. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A set of data for an advanced chemical exchange column has been obtained as R&D on the tritiated water processing. The packing and catalyst were uniformly mixed into the advanced column. It has been proved that there was possibility that the advanced column has a large separation factor in comparison with the ordinary columns, where the packing and catalyst were alternately packed. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column. The ion conductivity of the main material of the cell (Nafion) was hardly decreased even in the case of 1600 kGy irradiation by gamma ray.
Keywords: Tritium, Tritiated water, Permeation, Chemical exchange, Durability, Corrosion
Near-infrared spectroscopy of tritiated water K. Kobayashia, T. Enokidaa, D. Iioa, Y. Yamadaa,
M. Harab, Y. Hatanob
Abstract
It is quite important to control and observe the concentration and total amount of tritium when nuclear fusion is utilized like ITER project. There are many kinds of molecular species, conditions, amount, and concentration in the environment and it is desirable to have multiple ways of observation. Tritium is often found as hydrogen, water and methane molecules. Their behavior differs by the molecular species and detection of molecular species is therefore important. Near-infrared spectroscopy can be a good molecular species sensitive method for this purpose. However, since basic spectroscopic information of tritiated water (HTO or T2O) is unavailable, in this study we prepared tritiated water of high concentration, and carried out frequency modulation near-infrared spectroscopy. Measurement of highly tritiated water by imaging plate Y. Hatano1, M. Hara,1 H. Ohuchi,2 H. Nakamura3,
T. Hayashi,3 T. Yamanishi3
Abstract
Concentration of tritium in highly tritiated water was measured by exposing imaging plates (IPs) to water vapor. Tritium penetrated into photostimulated luminescence (PSL) phosphor through polyethylene terephthalate protection layer, and well detectable signal of PSL was induced at tritium concentration of 16 kBq cm-3. In addition, tritium was reversibly desorbed by keeping IPs in air, and signal from IPs returned to background level. In other words, IPs exposed to tritiated water vapor were reusable; tritium concentration in water could be measured without any waste. In addition, no handling of tritiated water such as sampling and dilution was necessary. Fusion Science and Technology, 60 (2011) 986-989 Y. Kawamuraa, W. Shub, M. MatsuyamaC, T. Yamanishia
Abstract
Beta ray induced X-ray spectrometry (BIXS) is one of the methods applicable to tritium gas monitor. It can measure tritium by counting the X-ray that is induced by interaction between the beta ray of tritium and the materials. Tritium gas monitor of BIXS use installed into Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency (JAEA) uses NaI(Tl) as the scintillator. In this work, the NaI scintillator and the photo-multiplier that can work at 150℃ have been installed instead of the ordinary scintillator and photo-multiplier. And, the sample gas such as He, T2, or T2 (1%)/He mixture was introduced into the tritium gas monitor kept at 120℃. Then, the counting rate was observed. The counting rate at 120℃ was about a half of that at the room temperature. The counting rate after the heating was almost same with that before the heating. So, the deterioration of the scintillator by the heating has not been observed. Fusion Science and Technology, 60 (2011) 1053-1056 R.-D. Penzhorn1, Y. Torikai1, M. Saito1, M. Hiro1,
A. Perevezentsev2, M. Matsuyama1
1Hydrogen Isotope Research Center, University of Toyama, 930 8555 Toyama, Gofuku 3190, Japan Abstract
The uptake of tritium on the surface and in the bulk of copper upon exposure to a 50 ℃ T/H mixture at 300 or 473 K was investigated using a chemical etching technique. Concentrations of tritium approaching saturation are achieved fairly rapidly in Cu even at low temperatures because of comparatively high diffusivity and low solubility of hydrogen in this material. The results were interpreted by a diffusion model. Most notorious are the very high concentrations of tritium on the topmost surface and subsurface. They were quantified by etching and confirmed by BIXS. In addition, there is evidence for tritium trapping in the subsurface region. Fusion Science and Technology, 60 (2011) 1057-1060 Y. Torikaia, M. Saitoa, A. Taguchia, R.-D. Penzhorna,
K. Akaishia, K. Tatenumab, K. Isobec, T. Hayashic,
T. Yamanishic
Abstract
A batch process concept for the decontamination from tritium of fusion reactor materials based on a hydrothermal treatment is under development at HRC. Essentially, tritium-loaded material is heated in a tightly closed vessel containing a defined amount of water. The objective of the water is to "capture" the released tritium in a small volume of liquid. For the detritiation, stainless steel temperatures in the range 393-473 K over a period of several days were found to be adequate. From the results it appears that by and large the released tritium accumulates in the purposely introduced water. The achieved degree of decontamination was estimated from the tritium concentration in the water and the tritium that remained in the decontaminated material. Tritium trapped in the surface layer of stainless steel was not reduced by the isochoric hydrothermal treatment in the same proportion as that in the bulk. Fusion Science and Technology, 60 (2011) 1061-1064 S. Fukadaa, Y. Edaoa, K. Satoa, T. Takeishia, K. Katayamaa,
K. Kobayashib, T. Hayashib, T. Yamanishib, Y. Hatanoc, A. Taguchic,
S. Akamaruc
Abstract
An experimental study on tritium transfer in porous concrete materials for the tertiary tritium safety containment is performed to investigate; (i) how fast tritium is transferred through porous concrete walls coated with or without a hydrophobic paint, and (ii) how well the hydrophobic paint coating works as a film protecting against tritium migrating through concrete. The experiment is comparatively carried out using two types of cement-paste and mortar disks with or without two kinds of paints. The results obtained here are summarized as follows: (1) Tritium transfer can be correlated in terms of the effective tritium diffusivity of DT=1.2x10-11 m2/s in porous cement. (2) Adsorbed or condensed liquid HTO itself is transferred only through pores in cement, and no tritium transfer path is present in non-porous sand. (3) Rates of tritium sorption and dissolution in cement and mortar coated with an epoxy-resin paint is correlated in terms of the diffusivity through the paint film of DT=1.0x10-16 m2/s. (4) The epoxy paint works more effectively as an anti-tritium diffusion coating than the acrylic-silicon resin paint. (5) The hydrophobic property of the silicon resin paint is deteriorated with elongating the contact time with H2O. Fusion Science and Technology, 60 (2011) 1155-1158 N. Nunomura1, S. Sunada2, K. Watanabe3
Abstract
Adsorption of H2O on the α-Al2O3 (0001) surface was studied by means of a first-principles calculation based on density functional theory (DFT). We also investigated the behavior of the isotope exchange by substituting a protium atom with deuterium or tritium. The oxygen atom of H2O adsorbs on the Al atom of the outermost surface layer, the entire water molecule is slanted at the direction of a hollow site, and a molecular plane is nearly parallel to the surface. The adsorbed states are mostly due to coupling of lone-pair electrons of H2O with the empty p orbitals of the Al atom of surface. The behavior of dissociation for H2O is clarified from molecular dynamics simulations, indicating that the second neighbor oxygen atom is more preferable adsorption site for dissociation than the nearest neighbor oxygen atom on the surface. Fusion Science and Technology, 60 (2011) 1395-1398 A. Taguchia, R. Akaia, M, Saitoa, Y. Torikaia,
M. Matsuyamaa, M. Ogurab, S. Uchidac
Abstract
The ability of various solid adsorbents to adsorb tritium from tritiated water was studied. The tritium removability and adsorption ability of mesoporous silica (MCM-41) were found to be larger than those of conventional microporous zeolites such as mordenite (MOR) and Linde-type A (LTA). The different adsorbents can be arranged in order of tritium removability and tritium adsorption ability as follows: MCM-41 > LTA(5A) > high-silica MOR [approximately equal] low-silica MOR [approximately equal] LTA(4A). The adsorbents can also be arranged in decreasing order of the separation factor (α) as follows: MCM-41 > LTA(5A) > low-silica MOR [approximately equal] LTA(4A) > high-silica MOR. Fusion Science and Technology, 60 (2011) 1455-1458 T. Odaa, M. Shimadab, K. Zhangc, P. Calderonib,
Y. Oyad, M .Sokolove, R. Kolasinskif, J. P. Sharpeb,Y. Hatanoc
Abstract
The behavior of hydrogen isotopes implanted into tungsten containing vacancies was simulated using a Monte Carlo technique. The correlations between the distribution of implanted deuterium and fluence, trap density and trap distribution were evaluated. Throughout the present study, qualitatively understandable results were obtained. In order to improve the precision of the model and obtain quantitatively reliable results, it is necessary to deal with the following subjects: (1) how to balance long-time irradiation processes with a rapid diffusion process, (2) how to prevent unrealistic accumulation of hydrogen, and (3) how to model the release of hydrogen forcibly loaded into a region where hydrogen densely exist already. Fusion Science and Technology, 60 (2011) 1155-1158 M. Saito, Y. Torikai, R.-D. Penzhorn, K. Akaishi, M. Matsuyama
Abstract
Uptake, distribution, and release behavior of tritium in Ni was investigated by chemical etching and thermal release rate measurements. Liberated tritium was found to consist almost exclusively of tritiated water. The chronic release rate of tritium from Ni was significantly larger than that from type 316 stainless steel. Depth profiles in specimens that partially lost tritium due to its chronic release into vacuum, air or a stream of argon could be reproduced by a one-dimensional diffusion model using best fit diffusion coefficients. Values of the best-fit diffusion coefficients at 298 K were found to be independent from the ambient into which tritium was released. The average diffusion coefficient from all measurements at 298 K, i.e. (2.7 ± 1.3) × 10-10 [cm2/s] was in line with diffusion coefficients calculated from literature data at the same temperature. Hence, the diffusion model constitutes a useful tool for the prediction of tritium bulk depth profiles in Ni during chronic release (CR). Fusion Science and Technology, 60 (2011) 1471-1474 Y. Togashi, M. Hara
Abstract
To understand the water vapor transport through a polypropylene film at near ambient temperature,
water vapor permeation and sorption measurements were carried out using tritiated water as tracer. The activation energy and frequency factor of the permeability were found to be 11 kJ/mol and 1.5 x 10-10 cm3(STP) cm cm-2 s-1 Pa-1, respectively. The corresponding values of the solubility were determined to be -30 kJ/mol and 2.9 x 10-10 cm3(STP) cm-3 Pa-1. Because the permeation can be described by a one-dimensional diffusion model, the diffusion coefficient was evaluated from the quotient of permeability and solubility. The activation energy of water diffusion through polypropylene was calculated to be 41 kJ/mol. Fusion Science and Technology, 60 (2011) 1491-1494 M. Matsuyamaa, K. Shinmuraa, Z. Chenb, Y. Torikaia
Abstract
Solubility of tritium in Cu-Be(2 mass%) alloy was determined by means of measurement of a tritium depth profile in the alloy. Tritium exposure to the samples was conducted under the following conditions: pressure, 0.4 to 2.6 kPa; temperature, 350 to 450℃ exposure time, 4 to 11 hours. Tritium depth profiles were obtained by chemical etching after the exposure. Remarkably high tritium concentration appeared in surface layers within 0.5 m, whereas almost constant concentration was observed from 10 m to the bulk. It was found, therefore, that surface tritium should be omitted in evaluation of the solubility of tritium. In addition, it was seen that dissolution of tritium into Cu-Be alloy obeys the Sieverts' law from the pressure dependence, and the solubility of tritium in Cu-Be alloy was lower than that in pure copper. From the temperature dependence of solubility, the heat of solution of tritium was determined as 17 kJ/mol. Fusion Science and Technology, 60 (2011) 1499-1502 Y. Ozeki, Y. Hatano, H. Taniguchi, M. Matsuyama
Abstract Fusion Science and Technology, 60 (2011) 1503-1506 K. Nishimuraa, K. Takatsukab, M. Matsuyamab, N. Nodab,
M. Tanakaa
Abstract
Hydrogen isotope retention and its removal in/from a plasma-facing wall and structural materials have been recognized as key issues for fusion reactor from the viewpoints of operational safety and environmental preservation. Baking and/or a glow discharge are possible effective methods to remove hydrogen isotopes from materials. To investigate the absorption and desorption characteristics of hydrogen isotopes in structural materials, a glow discharge apparatus with twin chambers (Glow-1, 2) made with stainless steel was provided. The sample tip can be moved between the Glow-1 chamber and Glow-2 chamber without exposing it to the air. The controlled infrared (IR) heating device was installed in the Glow-1 chamber to heat the sample tip. Results of absorption and desorption by glow discharge, and the thermal desorption by IR heating are analyzed. No oxides of hydrogen isotopes were observed in these experiments. Most gases are retained near the solid surface and are desorbed at a low temperature below 150℃ The activation energy was estimated to be about 50 kJ/mol. Fusion Science and Technology, 60 (2011) 1511-1514 S. K. Sharma1, H. Zushi2, I. Takagi3, Y. Hisano1,
T. Shikama3, S. Morita4, T. Tanabe1, N. Yoshida2, M. Sakamoto2,
Y. Higashizono2, K. Hanada2, M. Hasegawa2, O. Mitarai5,
K. Nakamura2, H. Idei2, K. N. Sato2, S. Kawasaki2, H. Nakashima2,
A. Higashijima2, Y. Nakashima6, N. Nishino3, Y. Hatano8,
A. Sagara4, Y. Nakamura4, N. Ashikawa4, T. Maekawa3, Y. Kishimoto3, Y.Takase9, QUEST2
Abstract
A permeation measuring system with a nickel membrane of 30 m thickness was installed near the mid plane of the spherical tokamak, QUEST. Hydrogen permeation through the membrane heated at fix temperatures (422 - 506 K) was measured during short pulse (< 1 s) and long pulse (1 hour) plasma discharges. After the membrane was heated to a required temperature, hydrogen plasma was discharged using a 2.45 GHz or 8.2 GHz RF system. Significant plasma-driven permeation was observed even for very short plasma discharges (e.g. 0.1s). Numerical calculations with the use of diffusion equation under recombination boundary conditions were conducted to simulate the transient permeation behavior. The numerical calculations were also used to estimate diffusion coefficient and recombination coefficients of membrane material. Temperature dependence of both the coefficients was explained by the Arrhenius law. A one hour long permeation curve was also numerically reproduced using the same set of parameters except an increasing recombination coefficient on the plasma side of the membrane. Fusion Science and Technology, 60 (2011) 1535-1538 Y. Nobutaa, Y. Yamauchia, T. Hinoa, S. Akamarub,
Y. Hatanob, M. Matsuyamab, S. Suzukic, M. Akibac
Abstract
Tritium retention in plasma facing materials is a primary issue for ITER and next step fusion devices, since it greatly affects its safety and operational schedule. In the ITER, carbon and tungsten are used as divertor materials. In the present study, co-deposited carbon film, tungsten and isotropic graphite were exposed to tritium gas, and then the amount of absorbed tritium was investigated. During the tritium exposure, the partial pressure of tritium gas was kept at 10 Pa. The sample temperature was kept a constant in the range from RT to 573 K. The amounts of absorbed tritium were evaluated by -ray-induced X-ray spectrometry (BIXS). The amounts of absorbed tritium in co-deposited carbon films were one or two orders of magnitude larger than that of polycrystalline tungsten and isotropic graphite. The amount of absorbed tritium for co-deposited carbon film with a high volume density (1.53 g/cm3) was several times larger than that of the film with a low volume density (1.13 g/cm3). The amount of absorbed tritium increased with the temperature. These results indicate that co-deposited carbon films can absorb much larger amount of tritium than tungsten and graphite, and carbon film density affects the amount of absorbed tritium. Journal of Nuclear Materials, 415(2011) S667-671 M. Shimadaa, Y. Hatanob, P. Calderonia, T. Odac,
Y. Oyad, M. Sokolove, K. Zhangb, G. Caof,
R. Kolasinskig, J.P. Sharpea
Abstract
With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites. Journal of Nuclear Materials, 415 (2011) S692-S695 N. Nunomura1, S. Sunada2, K. Watanabe3
Abstract
A permeation measuring system with a nickel membrane of 30 μm thickness was installed in a spherical tokamak QUEST. The membrane was located near the mid plane of the tokamak so that one side of the membrane was faced to the plasma. After the membrane was heated to 502 K, hydrogen plasma was discharged using 2.45 GHz RF system. The permeation flux through the membrane increased with a proper time-lag, that is, a significant plasma-driven permeation was observed. A numerical calculation of the diffusion equation under recombination boundary conditions was successfully conducted to simulate the transient permeation behavior. The recombination coefficients, estimated in a temperature range of 412- 575 K, can be explained by a model of thermally activated processes. Stable operation of the permeation system indicates the suitability of this system for the measurements of atomic flux with known diffusivity and recombination coefficients.
Journal of Nuclear Materials, 415 (2011) S701-S704 Y. Oyaa, M. Kobayashia, R. Kurataa,
W. Wanga, N. Ashikawab, A. Sagarab, N. Yoshidac,
Y. Hatanod, K. Okunoa
Abstract
To elucidate the simultaneous implantation effects on deuterium retention in polycrystalline tungsten, simultaneous 10 keV
C+, 3 keV D2+ and 3 keV He+ implantation was performed at room temperature. He+/D+ flux ratio dependence of deuterium retention indicated that He+ implantation induced high deuterium retention even if the He2+ flux ratio was low. D2 TDS spectrum for C+-D2+ implanted tungsten was clearly different from the samples with only D2+, D2+-He+ and C+-D2+- He+ implantation, indicating that D was trapped by C as C-D bond. For D2+ -He+ implantation sample, He+ implantation would introduce irradiation damages and D retention increased. However, C+- D2+-He+ implantation, the accumulation of C on tungsten was suppressed and the retention of D trapped by C clearly decreased.
These facts indicated the D retention for C+-D2+-He+ implanted tungsten was limited by He+ implantation and most of D would be retained in the interstitial sites or irradiation defects under simultaneous implantation with He.
Journal of Nuclear Materials, 415 (2011) S769-S772 T. Otsukaa, M. Shimadab, R. Kolasinskic, P. Calderonib,
J.P. Sharpeb, Y. Uedad, Y. Hatanoe, T. Tanabea
Abstract
We have applied a tritium imaging plate technique to measure the tritium distribution profile on surface and in bulk of various metal materials after exposure to a deuterium-tritium plasma in a linear plasma experimental apparatus. The experimental tritium concentration profiles in mm range are interpreted according to a simple hydrogen diffusion model in each metal. We found that a significant amount of tritium is localized in near surface regions and is clearly distinguishable from tritium diffused in the bulk. The amount of surface tritium is not likely correlated to bulk properties (diffusivity and solubility), but is related to trapping in surface defects or metal impurities such as oxide and carbide.
Journal of Nuclear Materials, 415 (2011) 555-558 W. Wang,a,e M. Kobayashia, R. Kurataa,
S. Suzukia, N. Ashikawab, A. Sagarab
N. Yoshidac, Y. Hatanod, G.N. Luoe, Y. Oyaa,
K. Okunoa
Abstract
The simultaneous implantation of 10 keV C+ and 3 keV D Journal of Nuclear Materials, 415 (2011) 1229-1232 K. Zhang, Y. Hatano
Abstract >
An electrolytic deposition technique was applied to seal open pores in sol-gel derived ZrO2 coating and to improve barrier effects against permeation of hydrogen isotopes. Disk-type specimens of type 430 ferritic stainless steel were first covered by thin ZrO2 films (50 nm) with a conventional sol-gel technique. Then, pores in the ZrO2 film was sealed with ZrO2 or Al2O3 by cathodic processes in ethanol solution of Zr or Al nitrate and subsequent heat treatments in air. The permeation rate of hydrogen was measured at 300-600 ℃. The sol-gel derived ZrO2 coatings showed only limited barrier effects; the permeation reduction factor (PRF) was about 6-800. Nevertheless, the treatments by electrolytic deposition technique resulted in considerable improvement in the barrier effects, especially at high temperature region (500℃), and the PRF increased to 100-1000.
Journal of Nuclear Materials, 415 (2011) S92-S95 Y. Uedaa, K. Miyataa, Y. Ohtsukaa, H.T. Leea,
M. Fukumotob, S. Brezinsekc, J.W. Coenenc, A. Kreterc,
A. Litnovskyc, V. Philippsc, B. Schweerc, G. Sergienkc,
T. Hiraid, A. Taguchie, Y. Torikaie, K. Sugiyamaf,
T. Tanabeg, S. Kajitah, N. Ohnoi, The TEXTOR team
Abstract W nanostructures (fuzz),
produced in the linear high plasma device, NAGDIS, were exposed to TEXTOR
edge plasmas (ohmic He/D mixed plasma and pure D plasma) to study formation, erosion and C deposition
on W fuzz in tokamak plasmas for the first time. Fuzz layers were either completely eroded or covered
by C deposit. There was no clear indication of W fuzz growth under the present conditions. There
was no significant difference of C deposition between 'thick' fuzz (500600 nm in thickness)
and 'thin' fuzz (300400 nm) in the He/D plasma. On the W fuzz surface,
C deposition was enhanced probably due to reduction of effective sputtering yield and effective reflection
coefficient of carbon ions, similar to roughness effects. Formation and erosion of W fuzz in tokamak
devices and role of impurities are discussed.
International Journal of Hydrogen Energy, 36 (2011) 7737-7746 V. N. Alimova, Y. Hatanob, A. O. Busnyuka, D. A. Livshitsc,
M.E. Notkina, A. I. Livshitsa
Abstract
The composite membranes based on Group 5 metals are capable of H2 separation with the high speed and infinite selectivity. The chemical and thermal stability are critical issues for the application of such membranes in the field of hydrogen energy. In order to understand the degradation mechanisms, the H2 permeation through composite Pd2 μm-Nb100μm-Pd2μm membranes was investigated in a very wide pressure range: (10-5-104) Pa. At higher pressures the surface contaminations only moderately decreased the permeation. However the permeation experiments at lower pressures demonstrated that an orders of magnitude change in the probability of H2 molecule dissociative sticking is actually hidden behind this relatively moderate effect. The membranes poisoned by the surface contaminations could be recovered by their exposure to O2 at (300-400) ℃. Heating at temperature higher than 500 ℃ resulted in the irreversible decrease of permeation and in the pronounced change of permeation behavior at the variation of H2 pressure. An extremely high permeation was observed at lower pressures at the clean surface of Pd coating. That allows developing an effective membrane pump for hydrogen isotopes.
Keywords: Hydrogen separation, Composite membranes, Niobium, Palladium
Fusion Science and Technology, 60 (2011) 288-291 H. Katsui1, A. Hasegawa2, Y. Katoh3,
Y. Hatano,4 T. Tanaka5, S. Nogami2, T. Hinoki2, T. Shikama1
Abstract
Compatibility of monolithic silicon carbide(SiC) with ternary lithium ceramics
(Li1-x AlO2-y, Li2-xTiO3-y,
Li2-xZrO3-y and Li4-xSiO4-y)
under irradiation of neutros at high temperatures was studied. Disk samples of SiC
in contact with sintered ternary lithium ceramics were irradiated in High Flux Isotope Reactor (HFIR)
at 800 ℃ to 5.9 displacements per atom(dpa).
Chemical reactions of SiC as determined by appearance of the surface
were relatively less significant for the systems of SiC/Li1-xAlO2-y and
SiC/Li2-x TiO3-y, whereas some bonding likely due to
chemical reaction between SiC and the lithium ceramics
and broken samples were observed in the systems of SiC/Li2-xZrO3-y
and SiC/Li4-xSiO4-y.
The effect of lithium burnup due to the (n, α) nuclear reaction was also
examined by using samples of lithium ceramics whose lithium ratio was
hypo-stoichiometric in the fabrication process. More reaction products were observed on the
surface of β-SiC in contact with Li1-xAlO2-y having the
lower lithium ratio (Li/Al). It was considered that the formation of LiAl5O8
phase due to lithim loss could deteriorate the compatibility of the SiC -Li1-x
AlO2-y system.
Surface Finising, 62 (2011) 94-98 T. Abe, M. Inoue
3190, Gofuku, Toyama-shi, Toyama 930-8555 No Abstract Keywords : Polygonal Barrel-Sputtering Method, Dry Process, Surface Modification, Zero-, One-, Three-Dimensional Materials, Functional Materials
Phys. Scr. T145 (2011) 014029-014034 Y. Ueda1, H. T. Lee1, N. Ohno2, S. Kajita3,
A. Kimura4, R. Kasada4,
T. Nagasaka5, Y. Hatano6,
A. Hasegawa7, H. Kurishita8, Y.Oya9
Abstract
The status of ongoing research projects of tungsten R&D in Japan is summarized in this paper.
For tungsten material development, a new improved fabrication technique, the so-called
superplasticity-based microstructural modification, is described. This technique successfully
improved fracture strength and ductility at room temperature. Recent results on vacuum
plasma spray W coating and W brazing on ferritic steels and vanadium alloys are explained.
Feasibility of these techniques for the manufacture of the blanket is successfully
demonstrated. The latest findings on the effect of neutron damage in tungsten on T retention
and on the change in mechanical and electrical properties are described. Retention
characteristics for neutron-damaged W were different compared to those for ion-damaged W.
Upon neutron irradiation, tungsten alloys containing transmutation elements of W (Re and Os)
show changes in properties that are different compared with those shown by pure W. The
effects of mixed plasma exposure (D/He/C) are described. Both D/He and D/C mixed ion
irradiations significantly affect ion-driven permeation in W. He bubble dynamics play a key
role in nano-structure formation on the W surface.
Phys. Scr. T145 (2011) 014029-014034 V. Kh. Alimov1,2, B. Tyburska-Püschel3, M. H. J't.Hoen4,
J. Roth3, Y. Hatano1, K. Isobe2, M. Matsuyama1, T. Yamanishi2
Abstract
Hydrogen isotope exchange in tungsten was investigated at various temperatures both after
sequential exposure to low-energy deuterium (D) and protium (H) plasmas and after
sequential irradiation with low-energy D and H ions. The methods used were thermal
desorption spectroscopy, and the D (3He, p) 4He nuclear reaction at 3
He energies varied from 0.69 to 4.0MeV, allowing the determination of the D concentration at depths of up to 6 μm.
It was found that a major portion of the deuterium initially accumulated in the D-implanted
W is released on subsequent exposure to H plasma or irradiation with H ions. Depth profiling
of D without and with subsequent H implantation shows strong replacement close to the
surface near room temperatures, but extending to all analyzable depths at elevated
temperatures.
>Phys. Scr. T145 (2011) 014044-014048 Y Hatano1, K Zhang1, K Hashizume2
Abstract
Layers of tetragonal ZrO2 (180 nm) were prepared on the surfaces of ferritic steel specimens
by dip coating and electrolytic deposition techniques. These methods were selected because of
their applicability to large and complicated structures. Hydrogen permeation tests were carried
out at 300-550 ℃ and the results were compared with those obtained in the previous study
for thinner ZrO2 coating (100 nm). Although the thickness of the present coatings was less
than twice that of the previous coatings, the permeation reduction factor for the former was
larger by an order of magnitude than that for the latter at/below 400 ℃. Tritium retention in
the coatings was also measured after exposure to a deuterium-tritium mixture at 300 ℃. The
obtained results suggest that the permeation rate was determined by transportation through
defects in the coatings, and the density of defects penetrating into the coating/substrate
interface was significantly reduced by an increase in coating thickness.
Phys. Scr. T145 (2011) 014050-014054 Yasuhisa Oya1, Masashi Shimada2, Makoto Kobayashi1,
Takuji Oda3, Masanori Hara4, Hideo Watanabe5, Yuji Hatano4,
Pattrick Calderoni2, Kenji Okuno1
Abstract
The behavior of D retention for Fe2+-irradiated tungsten with a damage of 0.025-3 dpa was
compared with that for neutron-irradiated tungsten with 0.025 dpa. The D2 thermal desorption
spectroscopy (TDS) spectra for Fe2+-irradiated tungsten consisted of two desorption stages at
450 and 550 K, while that for neutron-irradiated tungsten was composed of three stages and an
addition desorption stage was found at 750 K. The desorption rate of the major desorption
stage at 550K increased as the displacement damage increased due to Fe2+ irradiation
increasing. In addition, the first desorption stage at 450K was found only for damaged
samples. Therefore, the second stage would be based on intrinsic defects or vacancy produced
by Fe2+irradiation, and the first stage should be the accumulation of D in mono-vacancy and
the activation energy would be relatively reduced, where the dislocation loop and vacancy is
produced. The third one was found only for neutron irradiation, showing the D trapping by a
void or vacancy cluster, and the diffusion effect is also contributed to by the high full-width at
half-maximum of the TDS spectrum. Therefore, it can be said that the D2 TDS spectra for
Fe2+-irradiated tungsten cannot represent that for the neutron-irradiated one, indicating that
the deuterium trapping and desorption mechanism for neutron-irradiated tungsten is different
from that for the ion-irradiated one.
Phys. Scr. T145 (2011) 014051-014055 M. Shimada1, G. Cao2, Y. Hatano3, T. Oda4, Y. Oya4,
M. Hara3, P. Calderoni1
Abstract
Tungsten samples (99.99% purity from A.L.M.T. Corp., 6mm in diameter, 0.2mm in
thickness) were irradiated by high-flux neutrons at 50 ℃ C to 0.025 dpa in the High Flux Isotope
Reactor at Oak Ridge National Laboratory. Subsequently, the neutron-irradiated tungsten
samples were exposed to high-flux deuterium plasmas (ion flux: 1021-10
22m-2 s-1, ion fluence:
1025-1026 m-2)
in the Tritium Plasma Experiment at Idaho National Laboratory.
This paper reports the results of deuterium depth profiling in neutron-irradiated tungsten exposed to
plasmas at 100, 200 and 500 % via nuclear reaction analysis (NRA). The NRA measurements
show that a significant amount of deuterium (>0.1 at.% D/W) remains trapped in the bulk
material (up to 5μm) at 500 %. Tritium Migration Analysis Program simulation results using
the NRA profiles indicate that different trapping mechanisms exist for neutron-irradiated and
unirradiated tungsten.
Fusion Sci. Technol. 60 (2011) 321-328 M. Muroga1, D. K. Sze2, K. Okuno3,
T. Terai4, A. Kimura5, R. J. Kurtz6, A. Sagara1,
R. Nygren7, Y. Ued8, R. P. Doerner2, J. P. Sharpe9,
T. Kunugi10, N. B. Morley11, Y. Hatano12, M. A. Sokolov13,
T. Yamamoto14, A. Hasegawa15, Y. Katoh13, N. Ohno16,
K. Tokunaga17, S. Konishi5, S. Fukada17,
P. Calderoni9, T. Yokomine10, K. Messadek11, Y. Oya3,
N. Hashimoto18, T. Hinoki5,
H. Hashizume15, T. Norimatsu8, T. Shikama15,
R. E. Stoller13, K .A. Tanaka8, M. S. Tillack2
Abstract
Japan-US cooperation program TITAN (Tritium,
Irradiation and Thermofluid for America and Nippon)
started in April 2007 as 6-year project. This is the
summary report at the midterm of the project. Historical
overview of the Japan-US cooperation programs and
direction of the TITAN project in its second half are
presented in addition to the technical highlights.
Nucl. Fusion 51 (2011) 113020
Abstract
The behaviour of tungsten (W) plasma-facing components (PFCs) has been investigated in the plasma edge of the TEXTOR tokamak to study melt-layer ejection, macroscopic tungsten erosion from the melt layer as well as the changes of material properties such as grain-size and abundance of voids or bubbles. The parallel heat flux at the radial position of the exposed tungsten tile in the plasma ranges around q|| ~ 45 MW m-2 causing samples to be exposed at an impact angle of 35° to 20-30 MW m-2. Locally the temperature reached up to 6000 K, high levels of evaporation and boiling are causing significant erosion in the form of continuous fine spray or droplet ejection. The amount of fine-spray tungsten emission depends strongly on the material properties: in the case of the tungsten-tantalum alloy the effect of spraying and droplet emission is significantly higher at even low temperatures when compared with regular tungsten or even ultra-high purity tungsten which shows almost no spraying at all. Differences in the material composition, grain structure and size may be related to the different evolution of macroscopic erosion. In addition the re-solidified material is studied and strong differences in terms of re-crystallized grain size and evolution of the grain structure and grain orientation are observed. The build up of large voids has been observed. Proc. Radiochim. Acta 1, 49-53 (2011) H. Ohuchi1, Y. Hatano2
Abstract
Tritium measurement is indispensable for the
fuel-processing systems of deuterium-tritium (DT)-fusion facilities.
A new approach to detect tritium in regions deeper
than the escape depth of beta-rays from tritium is being
developed using an imaging plate (IP). The measurement
principle of this approach is to observe bremsstrahlung
X-rays induced by the tritium beta-rays. An IP made of
europium-doped BaFBr(I), a photostimulated luminescence
(PSL) material, is a two-dimensional radiation sensor. In
the present study, the characteristics of this IP for measuring
tritium by detecting bremsstrahlung X-rays, in particular
a fading effect and the energy dependence of PSL sensitivities,
are examined.
Electrochimica Acta 56 (2011) 8438-8445 C. Hiromi, M. Inoue, A. Taguchi, T. Abe
Abstract
The optimum Pt and Ru atomic composition of a carbon-supported Pt-Ru alloy (Pt-Ru/C) used in a practical direct methanol fuel cell (DMFC) anode was investigated. The samples were prepared by the polygonal barrel-sputtering method. Based on the physical properties of the prepared Pt-Ru/C samples, the Pt-Ru alloy was found to be deposited on a carbon support. The microscopic characterization showed that the deposited alloy forms nanoparticles, of which the atomic ratios of Pt and Ru (Pt:Ru ratios) are uniform and are in accordance with the overall Pt:Ru ratios of the samples. The formation of the Pt-Ru alloy is also supported by the electrochemical characterization. Based on these results, methanol oxidation on the Pt-Ru/C samples was measured by cyclic voltammetry and chronoamperometry. The results indicated that the methanol oxidation activities of the prepared samples depended on the Pt:Ru ratios, of which the optimum Pt:Ru ratio is 58:42 at.% at 25 ℃ and 50:50 at.% at 40 and 60 ℃. This temperature dependence of the optimum Pt:Ru ratio is well explained by the relationship between the methanol oxidation reaction process and the temperature, which is reflected in the rate-determining steps considered from the activation energies. It should be noted that at 25-60 ℃, the Pt-Ru/C with Pt:Ru = 50:50 at.% prepared by our sputtering method has the higher methanol oxidation activity than that of a commercially available sample with the identical overall Pt:Ru ratio. Consequently, the polygonal barrel-sputtering method is useful to prepare the practical DMFC anode catalysts with the high methanol oxidation activity. Key words: Direct methanol fuel cells (DMFCs), Pt-Ru/C, Pt and Ru atomic composition, Methanol oxidation, Polygonal barrel-sputtering method
Molecular Crystals and Liquid Crystals 538 (2011) 182-186 A. Aoki1, M. Naruse1, T. Abe2
Abstract
Series circuit of six organic thin-film solar cells (OSCs) consisting of poly(3-hexylthiophene) and PCBM has been
fabricated in order to electrolyze water into hydrogen and oxygen gases. The open circuit voltage increases
linearly with the number of unit cells and becomes 2.9 V at the six unit cells. On the other hand, the short circuit
current, the fill factor and the energy conversion efficiency are almost constant, independent of the number of unit
cells. The series circuit of six OSCs was combined with the water electrolysis cell with two platinum electrodes.
Hydrogen and oxygen gases were generated at each platinum electrodes of the electrolysis cell under illumination.
The operating current and voltage was determined to be 1.3 mA/cm2 and 2.6 V, respectively. Key words: Organic solar cells, Hydrogen gas, Energy conversion, Water electrolysis
It was found that the major deuterium desorption temperatures were located at around 660 K and
935 K, which correspond to the desorption of deuterium trapped as hydroxide. The deuterium retention
increased with increasing deuterium ion fluence, since the deuterium retention as hydroxide increased
significantly. However, retention saturated at an ion fluence of ~ 2.5 ×1021 D++ m-2.
The XPS result showed that FeOOD was formed on the surface, although pure Fe also remained in the oxide layer. These
facts indicate the nature of the oxide layer have a key role in deuterium trapping behavior.
[2011_02]
BIXS measurements of tritium uptake in C and W materials for EAST
b Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
[2011_03]
Pressure-composition isotherms of TiC1xH system at elevated temperatures
[2011_04]
Trapping and Depth Profile of Tritium in Surface Layers of Metallic Materials
[2011_05]
Alloying effects on the hydride formation of Zr(Mn1xCox)2
aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
b
Faculty of Human Development, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
[2011_06]
An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility
aFusion Safety Program, Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-7113, USA
bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
cRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
dInterdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581, Japan
eInstitute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011, Japan
[2011_07]
Comparison of hydrogen isotope retention and irradiation damage behaviors in tungsten and SS-316
with simultaneous C+-D2+ implantation
aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Oya,
Suruga-ku, Shizuoka 422-8529, Japan
bInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
cNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, Gifu 509-5292, Japan
dHydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japan
The total D retention for SS-316 with only D2+ implantation was about 45%
as large as that for tungsten. The D retention for simultaneous C+-D2+ implanted tungsten and SS-316 clearly increased as a factor of 1.7, which is almost the same among these samples. The density of dislocation loops was
enhanced by the simultaneous C+-D2+ implantation, indicating the D trapping site would be produced by C+ implantation. As for the D desorption temperature, small shift toward lower temperature side was found for SS-316 compared to tungsten, indicating the D trapping energy by dislocation loops and grain boundary for SS-316 is lower than that for tungsten.
[2011_08]
Recent activities of R&D on effects of tritium water on confinement materials and tritiated water processing
aJapan Atomic Energy Agency, Tokai, Ibaraki 319-1195 Japan
bUniversity of Toyama, Gofuku, Toyama 930-8555, Japan
cNagoya University, Chikusa-ku, Nagoya 464-8603, Japan
dShizuoka University, Ooya, Shizuoka 422-8529, Japan
[2011_09]
Fusion Science and Technology, 60 (2011) 941-943
aDepartment of Physics, University of Toyama, 3190 Gofuku, Toyama-city, Toyama, Japan, 930-8555
bHydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama-city, Toyama, Japan, 930-8555
The tritiated water was synthesized by the oxidation reaction of 1 Ci of T2. Near-infrared spectra at 1.3 micron were recorded. Many lines were observed which are not due to normal water. They are strong candidates of tritiated water spectral lines.
[2011_10]
Fusion Science and Technology, 60 (2011) 982-985
1Hydrogen Isotope Research Center, University of Toyama: Gofuku 3190, Toyama 930-8555, Japan
2Graduate School of Pharmaceutical Sciences, Tohoku University, Aramaki-Aoba, Aoba-ku, Sendai 980-8578, Japan
3Tritium Technology Group, Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195, Japan
[2011_11]
Stability of NaI (TI) detector for tritium monitor of BIXS use to hot environment
aJapan Atomic Energy Agency: Shirakata Shirane 2-4, Tokai, Ibaraki 319-1195, Japan
bITER organization, 13067 Saint-Paul-Lez-Durance Cedex, France
cUniversity of Toyama, Gofuku 3190, Toyama-city, Toyama 930-8555, Japan
[2011_12]
Uptake and distribution of tritium in copper
2ITER Organization, St. Paul-les-Durance, France
Tritium-loaded copper specimens release tritium chronically at ambient temperature. The egress of tritium manifests in the gas phase almost exclusively as tritiated water.
[2011_13]
Application of a hydrothermal treatment for the decontamination from tritium of fusion reactor materials
-Tritium decontamination using an autoclave-
aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
bKAKEN, Hori 1044, Mito, Ibaraki 310-0903, Japan
cTritium Technology group, Japan Atomic Energy Agency, Tokai-mura Naka-gun, Ibaraki, 319-1195, Japan
[2011_14]
Tritium transfer in porous concrete materials coated with hydrophobic paints
aDept. Advanced Energy Engineering Science, Kyushu University, Fukuoka 812-8581, Japan
b Tritium Process Laboratory of Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan
cHydrogen Isotope Research Center, University of Toyama, Gofuku, Toyama 930-8555, Japan
[2011_15]
First-principles study of water adsorption on α-Al2O3(0001):
influence of hydrogen isotope
1Information Technology Center Universiy of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
2Graduate School of Science and Engineering for Research, University of Toyama, 3190 Gofuku,
Toyama 930-8555, Japan
3Hydogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama 930-8555, Japan
[2011_16]
Tritium removal from tritiated water using mesoporous silica
aHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
bInstitute of Industrial Science, The Universiy of Tokyo, Komaba 4-6-1, Meguro-ku, Tokyo 153-8505, Japan
cGraduate School of Art and Science, The University of Tokyo, Komaba 3-811, Meguro-ku, Tokyo 153-8902, Japan
[2011_17]
Development of Monte Carlo simulation code to model behavior of hydrogen isotopes loaded into tungsten containing vacancies
aDepartment of Nuclear Engineering and Management, The University of Tokyo, Tokyo, 113-8656, JAPAN
bFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID, 83415, USA
cHydrogen Isotope Research Center, University of Toyama, Toyama, 930-8555, Japan
dRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka, 422-8529, JAPAN
eOak Ridge National Laboratories, Oak Ridge, TN, 37831, USA
fHydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, C4, 94551, USA
[2011_18]
On the fate of tritium in nickel
Hydogen Isotope Research Center, University of Toyama, 930 8555 Toyama, Gofuku 3190, Japan
[2011_19]
Water vapor permeability of polypropylene
Hydogen Isotope Research Center, Universy of Toyama: Gofuku 3190, Toyama, 930-8555, Japan
[2011_20]
Solubility of tritium in Cu-Be alloy
aHydogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
bSchool of Nuclear Science and Technology, University of Science and Technology of China Hefei, Anhui Province 230026, P. R. China
[2011_21]
Influence of chromium oxide layer on surface tritium contamination of type 316 stainless steel
Hydogen Isotope Research Center, University of Toyama: Gofuku 3190, Toyama 930-8555, Japan
[2011_22]
Absorption and desorption characteristics of hydrogen isotopes implanted into stainless steel
by glow discharge and baking
aNational Institute for Fusion Science, 322-6 0roshi-cho, Toki, 509-5292, Japan
bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama, 930-8555, Japan
[2011_23]
Hydrogen permeation measurements in the spherical tokamak QUEST and its numerical modeling
1IGSES, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan
2RIAM, Kyushu University, Kasuga, Fukuoka, 816-8580, Japan
3DNE, Graduate School of Engineering, Kyoto University, Japan
4National lnstitute for Fusion Science, Toki, Japan
5Kyushu Tokai University, 9-1-1 Toroku, Kumamoto 862-8652, Japan
6Plasma Research Center, University of Tsukuba, Japan
7DMSE, Graduate School of Engineering, Hiroshima University, Japan
8Hydrogen Isotope Research Center, Toyama University, Toyama 930-8555, Japan
9Graduate School of Frontier Science, University of Tokyo, Ibaragi, Japan
[2011_24]
Tritium absorption of co-deposited carbon films, graphite and polycrystalline tungsten
aLaboratoty of Plasma Physics and Engineering, Hokkaido University,
Kita-13, Nishi-8, Kita-ku, Sapporo 060-8628, Japan
bHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
cJapan Atomic Energy Agency, 801-1 Mukouyama, Naka-shi, Ibaraki, 311-0193 Japan
[2011_25]
First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE
aFusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
bHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
cDepartment of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656, Japan
dRadioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
eOak Ridge National Laboratory, Oak Ridge, TN 37831, USA
fDepartment of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706, USA
gHydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, CA 94551, USA
[2011_26]
Hydrogen permeation and recombination in Ni membrane placed on spherical tokamak QUEST
I. Takagia, S.K. Sharmab, H. Zushic,
R. Imadea, T. Komuraa, Y. Hisanob,
Y. Hatanod, Y. Nakamurae, A. Sagarae, N. Ashikawae
n
[2011_27]
Dynamics of hydrogen isotope trapping and detrapping for tungsten under simultaneous triple ion
(C+, D2+ and He+) implantation
aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836, Ohya, Suruga-ku, Shizuoka 422-8529, Japan
bNational Institute for Fusion Science, 322-6 Oroshi-cho, Toki-shi, Gifu 509-5292, Japan
cInstitute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
dHydrogen Isotope Research Center, University of Toyama, 3190, Gofuku, Toyama 930-8555, Japan
[2011_28]
Application of tritium imaging plate technique to examine tritium behaviors on the surface and in the bulk of plasma-exposed materials
aInterdisciplinary Graduate School of Engineering and Science, Kyushu University, Japan
bFusion Safety Program, Idaho National Laboratory, USA
cHydrogen and Metallurgical Science Department, Sandia National Laboratories, USA
dGraduate School of Engineering, Osaka University, Japan
eHydrogen Isotope Research Center, University of Toyama, Japan
[2011_29]
Temperature dependence of retention of energetic deuterium and carbon simultaneously implanted into tungsten
aRadioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529, Japan
bNational Institute for Fusion Science, Gifu, Japan
cResearch Institute for Applied Mechanics, Kyushu University, Kyushu, Japan
dHydrogen Isotope Research Center, University of Toyama, Toyama, Japan
eInstitute of Plasma Physics, Chinese Academy of Sciences, Hefei, China
into tungsten was carried out at elevated temperatures to establish the retention mechanism of energetic deuterium and carbon. Thermal desorption spectroscopy showed that the deuterium retention obviously decreased as the implantation temperature increased. All the desorption stages disappeared when the implantation was performed at 673 K, indicating no deuterium trapping as C-D bonds was processed in this temperature. The results of X-ray photoelectron spectroscopy and glow discharge-optical emission spectroscopy showed that a mixed carbon layer had been formed during the implantation, resulting from the carbon deposition and accumulation near the surface of tungsten during the implantation.
The carbon layer would enhance the chemical sputtering with deuterium and reduce the deuterium retention, as the implantation temperature increases. During discussion, a simple retention mechanism has been proposed, which shows the importance of implantation temperature.
[2011_30]
Sealing of pores in sol-gel-derived tritium permeation barrier coating by electrochemical technique
Hydrogen Isotope Research Center, University of Toyama, Toyama, Japan
[2011_31]
Exposure of tungsten nano-structure to TEXTOR edge plasma
aGraduate School of Engineering, Osaka University, Suita, Osaka 565-0871, Japan
bJapan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193, Japan
cInstitute für Energieforschung, Forschungszentrum Julich, Association EURATOM-FZJ, Germany
dITER Organization, 13067 St. Paul-lez-Durance, France
eHydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555, Japan
fMaxPlanckInstitut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
gInterdisciplinary Graduate School of Engineering Science, Kyushu University, Fukuoka, Japan
hEcoTopia Science Institute, Nagoya University, Nagoya 464-8603, Japan
iGraduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan
[2011_32]
Hydrogen permeation through the Pd-Nb-Pd composite membrane: Surface effects and thermal degradation
aBonch-Bruevich University, 61 Moika, St. Petersburg 191186, Russia
bHydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
cInnolume GmbH, Konrad-Adenauer-Allee 11, 44263 Dortmund, Germany
[2011_33]
Study on compatibility between silicon carbide and solid breeding materials under neutron irradiation
1Institute for Materials Research, Tohoku University, Sendai, Japan
2Department of Quantum Science and Energy Engineering, Tohoku University Sendai, Japan
3Malerials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN
4Hydrogen lsolope Research Center, University of Toyama, Toyama, Japan
5Department of Helical Plasma Research, National Institute for Fusion Science, Toki, Japan
6Institute of Advanced Energy, Kyoto University, Kyoto, Japan
[2011_34]
*This paper is written in Japanese.
Novel surface modification process by the polygonal barrel-sputtering method
Hydrogen Isotope Research Center, University of Toyama
[2011_35]
Recent progress of tungsten R&D for fusion application in Japan
1Graduate School of Engineering, Osaka University, Osaka 565-0871, Japan
2Graduate School of Engineering, Nagoya University, Nagoya 464-8603, Japan
3EcoTopia Science Institute, Nagoya University, Nagoya 464-8603, Japan
4Institute of Advanced Energy, Kyoto University, Kyoto 611-0011, Japan
5National Institute for Fusion Science, Gifu 509-5292, Japan
6Hydrogen Isotope Research Center, Toyama University, Toyama 930-8555, Japan
7Department of Quantum Science and Energy Engineering, Tohoku University, Sendai 980-8579, Japan
8Institute for Materials Research (IMR), Tohoku University, Ibaraki 311-1313, Japan
9Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
[2011_36]
Hydrogen isotope exchange in tungsten irradiated sequentially with low-energy deuterium and protium ions
1Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
2Tritium Technology Group, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan
3Max-Planck-Institut fur Plasmaphysik, EURATOM Association, D-85748 Garching, Germany
4Institute for Plasma Physics Rijnhuizen, EURATOM-FOM, NL-3439 MN Nieuwegein, The Netherlands
[2011_37]
Fabrication of ZrO2 coatings on ferritic steel by wet-chemical methods as a
tritium permeation barrier
1Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
2Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, Fukuoka 812-8581,
Japan
[2011_38]
Comparison of deuterium retention for ion-irradiated and neutron-irradiated tungsten
1Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529, Japan
2Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
3Department of Nuclear Engineering and Management, School of Engineering, The University of
Tokyo, Tokyo 113-8656, Japan
4Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
5Institute for Applied Mechanics, Kyushu University, Kasuga, Fukuoka 816-8580, Japan
[2011_39]
The deuterium depth profile in neutron-irradiated tungsten exposed to plasma
1 Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415, USA
2Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706, USA
3 Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555, Japan
4Department of Nuclear Engineering and Management, The University of Tokyo,
Tokyo 113-8656, Japan
[2011_40]
Midterm summary of Japan-US fusion cooperation program TITAN
1NIFS, Toki, Japan, 2UCSD, San-Diego, CA, USA, 3Shizuoka University, Shizuoka, Japan,
4University of Tokyo, Tokyo, Japan, 5Kyoto University, Uji, Japan, 6PNNL, Richland, WA, USA,
7SNL, Albuquerque, NM, USA, 8Osaka University, Suita, Japan, 9INL, Idaho Falls, ID, USA,
10Kyoto University, Kyoto, Japan, 11UCLA, Los Angeles, CA, USA,
12Toyama University, Toyama, Japan,
13ORNL, Oak Ridge, TN, USA, 14UCSB Santa-Barbara, CA, USA,
15Tohoku University, Sendai, Japan, 16Nagoya University, Nagoya, Japan,
17Kyushu University, Kasuga, Japan, 18Hokkaido University, Sapporo, Japan
[2011_41]
Melt-layer ejection and material changes of three different tungsten materials under high heat-flux conditions in the tokamak edge plasma of TEXTOR
1Institute for Energy and Climate Research, Forschungszentrum Julich, EURATOM Association,Trilateral Euregio Cluster, Julich, Germany, 2SCK-CEN, Belgian Nuclear Research Centre, Association EURATOM, Partner In the Trilateral Euregio Cluster, Mol, Belgium, 3Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany
4Hydrogen Isotope Research Center, University of Toyama, Japan, 5International Research Center for Nuclear Materials Science, Institute for Materials Research (IMR), Tohoku University, Oarai, Ibaraki, Japan,
6Graduate School of Engineering, Osaka University, Japan
[2011_42]
Tritium measurement using a photo-stimulable phosphor BaFBr(I):Eu2+plate
1Graduate School of Pharmaceutical Sciences, Tohoku University, 6-3 Aoba, Aramaki, Aoba-ku, Sendai, Miyagi, 980-8578 Japan
2Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, Toyama, 930-8555 Japan
[2011_43]
Optimum Pt and Ru atomic composition of carbon-supported Pt-Ru alloy electrocatalyst for methanol oxidation studied by the polygonal barrel-sputtering method
Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, Toyama, 930-8555 Japan
[2011_44]
Energy Conversion into Hydrogen Gas Using Series Circuit of Organic Thin-film Solar Cells
1Materials Science & Engineering, Graduate School of Engineering, Nagoya Institute of Technology, Gokiso, Showa-ku, Nagoya 466-8555, Japan
2Hydrogen Isotope Research Center, University of Toyama, 3190 Gofuku, Toyama, Toyama 930-8555, Japan