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Journal Articles

Behavior of environmental tritium at NIFS Toki Site of Japan

Sugihara, Shinji*; Tanaka, Masahiro*; Tamari, Toshiya*; Shimada, Jun*; Takahashi, Tomoyuki*; Momoshima, Noriyuki*; Fukutani, Satoshi*; Atarashi-Andoh, Mariko; Sakuma, Yoichi*; Yokoyama, Sumi*; et al.

Fusion Science and Technology, 60(4), p.1300 - 1303, 2011/11

 Times Cited Count:1 Percentile:11.72(Nuclear Science & Technology)

The purpose of this study is to develop the technique to evaluate the environmental tritium behavior of the nuclear facility origin. Tritium concentrations of river water, precipitation and ground water around the NIFS site were determined by low background liquid scintillation measurement system combined with the electrolysis using solid polymer electrolyte. The electric conductivity and flow rate of the river and isotopic ratio of oxygen and hydrogen of water samples were also measured. The tritium concentrations in precipitation showed the seasonal variation and the range were 0.09-0.78 Bq/L. The tritium concentrations of river water and ground water were almost constant, 0.34 and 0.24 Bq/L respectively. The simple dynamic model for the site around the NIFS facilities was developed using measured data, and the behavior of tritium was simulated.

Journal Articles

Detritiation behavior of HTO in a epoxy paint

Kobayashi, Kazuhiro; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1335 - 1338, 2011/11

 Times Cited Count:1 Percentile:11.72(Nuclear Science & Technology)

In a fusion reactor of high safety and acceptability, safety confinement of tritium is one of key issues for the fusion reactor. Tritium should be well-controlled and not excessively released to environment and to prevent workers from excess exposure. Especially, the hot cell and tritium facility of ITER will be used various construction materials such as the concrete and the organic materials. Transport properties of tritiated water vapor (HTO) in the epoxy paint has been evaluated by the HTO exposure and removal behavior from the epoxy paint in order to obtain the data base of tritium behavior in the confinement facilities such as the hot cell or the tritium plant building of ITER.

Journal Articles

Measurements of carbon dust property in experiment and post-campaign sampling on JT-60U Tokamak

Asakura, Nobuyuki; Hayashi, Takao; Ashikawa, Naoko*; Hatae, Takaki; Nakano, Tomohide

Fusion Science and Technology, 60(4), p.1572 - 1575, 2011/11

 Times Cited Count:5 Percentile:42.62(Nuclear Science & Technology)

Dust research has been performed in JT-60U in order to predict the plasma performance and the tritium retention in a fusion reactor. Laser scattering measurement showed, in specific discharge after disruptions, both the size and number were peaked in the far-SOL and they decreased near the separatrix. This result shows that sublimation of dust is dominant in the SOL. Dust collection after the experiment campaign showed that large weight of the dust was cumulated on the exhaust route of gas flow under the divertor. The total amount of the hydrogen isotope contained in the dust were estimated for the cases with deposited in the volume and near the surface.

Journal Articles

Measurement of dust quantity and distribution collected from JT-60U

Hayashi, Takao; Asakura, Nobuyuki; Ashikawa, Naoko*; Nakano, Tomohide

Fusion Science and Technology, 60(4), p.1548 - 1551, 2011/11

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

A real mass densities of carbon dust collected in the baffle and divertor regions of JT-60U were investigated. On the plasma-exposed surface, large areal density of 610 mg/m$$^{2}$$ is found at the upper tile of the inner divertor, which is much larger than other areas due to the soft deposition. On the other hand, as for the plasma-shadowed area, largest areal density of 5,100 mg/m$$^{2}$$ was found underneath the dome structure. The total dust weights at the plasma-exposed surface and the shadowed areas were estimated to be 1.3 g and 22.2 g, respectively, assuming the toroidal symmetry. Count-based size distributions were also investigated. The average dust size of the main population is less than 20 $$mu$$m for both the plasma-exposed surface and the shadowed area.

Journal Articles

Effects of tritiated water on corrosion behavior of SUS304

Oyaizu, Makoto; Isobe, Kanetsugu; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1515 - 1518, 2011/11

 Times Cited Count:3 Percentile:28.8(Nuclear Science & Technology)

The effects of tritiated water on the corrosion behavior of SUS304 stainless steel was studied using Tafel extrapolation method, one of electrochemical techniques with changing tritium concentration, dissolved oxygen concentration and pH in electrolyte as parameters. It was indicated that there would be two or more effects of tritium that enhance the corrosion of SUS304 stainless steel under several experimental conditions. One is passivation inhibitory effect, which could be observed only in highly corrosive circumstance of 1N H$$_{2}$$SO$$_{4}$$ electrolyte. The other effects of tritium on corrosion behavior could be observed not only in 1N H$$_{2}$$SO$$_{4}$$ but also in corrosive circumstance of 1N Na$$_{2}$$SO$$_{4}$$ electrolyte, which would be affected by dissolved oxygen concentration as well as tritium concentration.

Journal Articles

Development of high efficiency electrode for highly tritiated water processing

Isobe, Kanetsugu; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1387 - 1390, 2011/11

 Times Cited Count:1 Percentile:11.72(Nuclear Science & Technology)

Aiming to enhance the efficiency of ceramic electrolysis method, we developed new electrodes using cerium oxide (Ceria). We prepared electrodes by two manufacturing methods. One is to mix ceria into Pt paste and then electrode was sintering on the YSZ. The other is to use Ceria as intermediate layer between YSZ and Pt-YSZ electrode. The water decomposition performance of such electrodes and usual electrode using Pt-YSZ was confirmed in different humidity at 1073 K. Both electrodes using Ceria showed higher water decomposition performance than that of usual electrode. Especially, 30% ceria adding electrode showed highest performance and the decomposition efficiency was one order magnitude higher than that of usual electrode.

Journal Articles

Radiochemical reactions between tritium and carbon dioxide at elevated temperatures

Isobe, Kanetsugu; Nakamura, Hirofumi; Nakamichi, Masaru; Yamanishi, Toshihiko

Fusion Science and Technology, 60(4), p.1584 - 1587, 2011/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

We focused on the reaction between tritium and carbon dioxide at elevated temperature, self-radiation reactions at 373, 473 and 573 K were investigated. Self-radiation experiment using high purity gaseous tritium was carried out in the chamber of stainless steel at atmospheric pressure and initial ratio between gaseous tritium and carbon dioxide was almost 1:1. After 2 weeks experiment, gas contents and these concentrations were measured with quadrupole mass spectrometer. Main products were carbon monoxide, water and methane. And production ratios of such products were almost same at any temperature. Therefore, it was found that self-radiation reaction between gaseous tritium and carbon dioxide is independent of temperature in the rage of 373-573 K.

Journal Articles

HTO contamination on polymeric materials

Iwai, Yasunori; Kobayashi, Kazuhiro; Yamanishi, Toshihiko

Fusion Science and Technology, 60(3), p.1025 - 1028, 2011/10

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

We have tested a number of polymeric materials used as gasket, insulator, glove and casing panel in the solid-polymer-electrolyte (SPE) tritiated water electrolyzer to evaluate the contamination by tritiated water and the change in contamination by irradiation. HTO contamination on polymeric materials both being exposed to 740-1110 Bq/cm$$^{3}$$ of HTO vapor with a 1kPa of H$$_{2}$$O pressure and being immersed in 70000 Bq/cm$$^{3}$$ of HTO water was considered in the test. The exposed time affected negligibly the total amount of leached HTO from the rubber samples exposed to HTO vapor. The immersed time in contrast affected strongly the total amount of leached HTO from the rubber samples. The total amount of leached HTO from radiation-crosslinkable butyl rubber and radiation-degradable perfluoro Karlez rubber immersed in HTO was considerably increased as the integrated dose was increased. However, we found that the total amount of leached HTO from the irradiated rubber can maintain the similar amount from unirradiated by setting the hydrogen/fluoride ratio of the polymeric component to the suitable number.

Journal Articles

Study of tritium and helium release from irradiated lithium ceramics Li$$_{2}$$TiO$$_{3}$$

Kulsartov, T.*; Tazhibayeva, I.*; Gordienko, Y.*; Chikhray, E.*; Tsuchiya, Kunihiko; Kawamura, Hiroshi; Kulsartova, A.*

Fusion Science and Technology, 60(3), p.1139 - 1142, 2011/10

 Times Cited Count:10 Percentile:65.84(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Behavior of tritiated water on concrete materials

Kobayashi, Kazuhiro; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 60(3), p.1041 - 1044, 2011/10

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

In a fusion reactor of high safety and acceptability, safety confinement of tritium is one of key issues for the fusion reactor. Tritium should be well-controlled and not excessively released to environment and to prevent workers from excess exposure. Especially, the hot cell and tritium facility of ITER will be used various construction materials such as the concrete and the organic materials. Since the organic and the concrete materials will be contaminated by tritium compared with the metal materials such as SS, it is very important to study the tritium behavior on the materials from viewpoint of protection the excess exposure to workers. Therefore, in order to understand for tritium behavior on the concrete materials, the sorption and desorption experiment was carried out as a function of the exposure time, temperature and tritiated water concentration. From the results, the behavior of tritium sorption and desorption in the concrete materials will be discussed.

Journal Articles

Past 25 years results for large amount of tritium handling technology in JAEA

Yamanishi, Toshihiko; Yamada, Masayuki; Suzuki, Takumi; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Inomiya, Hiroshi; Hayashi, Takumi

Fusion Science and Technology, 60(3), p.1083 - 1087, 2011/10

 Times Cited Count:2 Percentile:20.43(Nuclear Science & Technology)

Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency has been established as the only test facilities to handle over 1 gram of in Japan. From March 1988, TPL has been operated with tritium, and no tritium release accident has been observed. The average tritium concentration in a stream from a stack of the TPL to environment was 71 Bq/m$$^{3}$$, and was 1/70 of the Japanese regulation value for HTO. The failure data have been analyzed for several main components of the safety systems such as pumps, valves, and monitors. The data on the tritium waste and accountancy has also been accumulated. As a study of the Grants-in-Aid for Scientific Research, these data are analysed and are reported.

Journal Articles

Stability of Nal(Tl) detector for tritium monitor of BIXS use to hot environment

Kawamura, Yoshinori; Shu, Wataru*; Matsuyama, Masao*; Yamanishi, Toshihiko

Fusion Science and Technology, 60(3), p.986 - 989, 2011/10

 Times Cited Count:4 Percentile:36.36(Nuclear Science & Technology)

Assuming the blanket sweep gas at the outlet of the blanket, tritium gas monitor by $$beta$$-ray induced X-ray spectroscopy has been modified, and has measured tritium at 120 $$^{circ}$$C. The counting rate at 120 $$^{circ}$$C was about 1/2 of that at the room temperature. In this work, the measurement system was a closed system. When two systems have same volume and same pressure, the number of molecules in higher temperature system is smaller. This is one of the causes of small counting rate. The deterioration of the scintillator after heating was not observed.

Journal Articles

Improvement of ultimate pressure of oil-free reciprocating pump for tritium service

Hayashi, Takumi; Yamada, Masayuki; Suzuki, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 60(3), p.1101 - 1104, 2011/10

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Oral presentation

Experimental study on helium gas flow through pebble bed of ceramic tritium breeder in a test blanket module

Yoshikawa, Akira; Seki, Yohji; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi; Enoeda, Mikio; Fukada, Satoshi*

no journal, , 

The distributions of the helium gas flow in the pebble beds were estimated by the measurement and the calculation of the pressure drops from the view point of the effective tritium recovery in a Japanese test blanket module for ITER. The pebbles of a glass or Li$$_{2}$$TiO$$_{3}$$ with the diameter of 1.0 mm were packed and the helium gas was purged with the flow rates of 0-50 L/min in the membrane box. The pressure drops on the flow rates of the helium gas was measured with the differential manometer in each pebble bed. It was observed that the distribution of the pressure drop was linearly increased with the increase of the flow rate in all the cases of the pebble beds. The pressure drop in Li$$_{2}$$TiO$$_{3}$$ pebble bed was larger than that of the glass pebble bed. It was considered that the difference of the pressure drops between the different kinds of the pebbles was derived from the difference of the surface roughness of pebbles.

Oral presentation

Recent activities on tritium technologies of BA DEMO-R&D program

Yamanishi, Toshihiko; Hayashi, Takumi; Yamada, Masayuki; Kobayashi, Kazuhiro; Nozawa, Takashi; Ohira, Shigeru; Iwai, Yasunori

no journal, , 

A RI handling equipment has been designed, and manufactured at Rokkasho in Japan. The equipment is the first and quite unique facility in Japan, where tritium, beta and $$gamma$$ RI species, and beryllium can be used simultaneously. The safety analysis of the facility has been carried out for the licensing. Both the external and internal radiation dose for workers was quite smaller than the regulation values. Some preliminary studies have been started with Japanese Universities for the tritium analysis and for the basic tritium safety at Tritium Process Laboratory and Japanese Universities in cooperation with the works related to the Grants-in-Aid for Scientific Research.

Oral presentation

Tritium distribution of tungsten exposed with low energy, high flux D plasma

Isobe, Kanetsugu; Alimov, V.; Yamanishi, Toshihiko; Torikai, Yuji*

no journal, , 

To understand these plasma surface interactions, tritium distribution of tungsten exposed with low energy (38 eV), high flux D plasma was examined with BIXS. D plasma exposures were carried out at around 495 and 550 K of specimen. After that, specimen was exposed with gaseous tritium diluted with deuterium at 473 K in 5 hours. Amount of tritium in surface layer was measured with BIXS and tritium distribution of surface. The amount of tritium in surface layer was different of each exposure condition and tungsten exposed at 495 K shows highest amount of tritium. This result quite agrees with D inventory examination with thermal desorption spectrometer.

Oral presentation

Tritium removal test for decommissioning of FUGEN

Matsushima, Akira; Ishiyama, Masahiro; Matsuo, Hidehiko; Sato, Yuji

no journal, , 

no abstracts in English

Oral presentation

Present status of the Broader Approach activities in Japan

Okumura, Yoshikazu

no journal, , 

The Broader Approach (BA) activities aim at complementing the ITER project and at an early realization of fusion energy by carrying out R&D and developing some advanced technologies for the future demonstration power reactor (DEMO). They consist of (1) Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) project, (2) International Fusion Energy Research Center (IFERC) project, and (3) Satellite Tokamak Program. A new research center named International Fusion Energy Research Center has been established in Rokkasho, Aomori. The construction of the research buildings has been completed in March 2010. Installation of the experimental equipments for the DEMO R&D, and the accelerator components for the IFMIF/EVEDA project are under fabrication. A high performance computer with 1 Peta flops will be opertional in January 2012 for the simulation of the burning plasma, etc.

Oral presentation

Hydrogen isotope retention on dust particles in fusion devices

Ashikawa, Naoko*; Asakura, Nobuyuki; Hayashi, Takao; Fukumoto, Masakatsu; Kurata, Rie*; Kobayashi, Makoto*; Muroga, Takeo*; Oya, Yasuhisa*; Okuno, Kenji*

no journal, , 

Dust control is an important issue related to tritium retentions in thermonuclear fusion devices with magnetic confinements. In particular, remained hydrogen isotope in dust particles at shadow areas are serious problem related the limitation t capacity in ITER. In previous work of dust collection in LHD and JT-60U, typical diameters of dust particles were shown to be 0.001-10 micron in LHD, and 0.1-100 micron in JT-60U. Some tokamak device groups in TFTR, TEXTOR, JT-60 and JET tried to measure the hydrogen isotopes retentions in dust particles. However, they measured the retention only in the dust flakes with sizes exceeding about 10 microns. In this study, we tried to measure hydrogen isotope retentions in the dusts with small amount including those with small diameters. The estimation of the amount of remained hydrogen isotopes in the dust particles will be useful to estimate tritium inventory and optimize cleaning methods specific dusts in ITER and future fusion devices.

Oral presentation

Application of HTO as radioactive tracer to investigate water transport properties of fluoropolymer-based fuel-cell electrolyte membranes

Sekine, Toshihiko; Sawada, Shinichi; Yamaki, Tetsuya; Asano, Masaharu; Suzuki, Akihiro*; Terai, Takayuki*; Maekawa, Yasunari

no journal, , 

A tracer permeation technique was applied as the only way to detect water transport in polymer electrolyte membranes (PEMs) immersed in a water/methanol mixture. As a first step, we investigated the applicability of a tritiated water (HTO) tracer method comparing with the results from a heavy oxygen water (H$$_{2}$$$$^{18}$$O) tracer. According to a simple isotopic-exchange theory, the measured HTO permeation coefficient, P(HTO), is assumed to depend on the abundance ratio between H$$_{2}$$O and CH$$_{3}$$OH in the solution used for the permeation experiments. This assumption enabled us to estimate the effective permeability of HTO, P(HTO)$$_{eff}$$. Importantly, we noticed a rather small difference between the P(HTO)$$_{eff}$$ and P(H$$_{2}$$$$^{18}$$O) values, which probably indicates that HTO mobility was affected by the different T/H substitution rates due to hydrogen-bonding interaction with polar sulfonic acid groups in PEMs.

Oral presentation

Performance of new tritium calorimeter in TPL/JAEA

Hayashi, Takumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Suzuki, Takumi; Yamada, Masayuki; Nakamura, Hirofumi; Yamanishi, Toshihiko

no journal, , 

21 (Records 1-20 displayed on this page)