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Journal Articles

Surface analysis for the TFTR Armor tile exposed to D-T plasmas using nuclear technique

Kubota, Naoyoshi; Ochiai, Kentaro; Kutsukake, Chuzo; Hayashi, Takao; Shu, Wataru; Kondo, Keitaro; Verzilov, Y.*; Sato, Satoshi; Yamauchi, Michinori; Nishi, Masataka; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 7 Pages, 2007/03

Fuel and impurity particles show complicated behavior on the surface of plasma facing components (PFC) in fusion devices. The study is important for the design of the fuel recycling, safety management of the tritium inventory, etc. Quantitative measurements of hydrogen and lithium isotopes together with other impurities on the PFC surface exposed to D-T plasmas in TFTR were performed using the deuteron-induced nuclear reaction analysis, imaging plate method, full combustion method and activation analysis. The tritium depth profile was different from deuterium one. The surface tritium largely contributed to the whole tritium in the sample. On the other hand, the retained amount of lithium-6 was lager than that of lithium-7. This relates to the injection of enriched lithium-6 pellets in some campaigns. No other impurities were detected. So the large amount of tritium remained near the surface and did not diffuse more deeply, which gives a bright prospect for tritium safety.

Journal Articles

Safety handling characteristics of high-level tritiated water

Hayashi, Takumi; Ito, Takeshi*; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Nishi, Masataka

Fusion Engineering and Design, 81(8-14), p.1365 - 1369, 2006/03

 Times Cited Count:18 Percentile:19.87(Nuclear Science & Technology)

In a fusion reactor, high-level tritiated water of more than GBq/ml will be generated and stored temporally in the various areas. High level tritiated water decomposes by itself and generates hydrogen and oxygen, and becomes to tritiated hydrogen peroxide water, however, effective G-values from tritiated water are different from those obtained $$gamma$$-ray experiments in our previous report. Furthermore, tritiated water of about 250GBq/ml has been stored for several years safely and checked its characteristics. Using the above experiences, this paper summarizes safety requirements for storage of high-level tritiated water and discusses design issues of the safety storage system. Concerning gaseous species, storage tank should be maintained at negative pressure and purged periodically or constantly to dedicated tritium removal system. Specially, it is important that the G-value of high-level tritiated water is increasing with decreasing the tritium concentration. The pH and ORP (Oxidation Reduction Potential) of tritiated water have been also changed depending on the tritium concentration and maintained for more than several years in glass vessel. High-level tritiated water of more than GBq/ml was acid and became to be corrosive depending on the dissolved species. Large amount of tritiated water will be stored in the various tanks of stainless steel, therefore, it should be monitored so that the liquid situation is maintained not to be corrosive.

Journal Articles

Distinctive radiation durability of an ion exchange membrane in the SPE water electrolyzer for the ITER water detritiation system

Iwai, Yasunori; Yamanishi, Toshihiko; Isobe, Kanetsugu; Nishi, Masataka; Yagi, Toshiaki; Tamada, Masao

Fusion Engineering and Design, 81(1-7), p.815 - 820, 2006/02

 Times Cited Count:14 Percentile:27.2(Nuclear Science & Technology)

Solid-polymer-electrolyte (SPE) water electrolysis is attractive in electrolytic process of water detritiation system (WDS) in fusion reactors because it can electrolyze liquid waste directly, but radioactive durability of its ion exchange membrane is a key point. Radioactive durability of Nafion, a typical commercial ion exchange membrane, was experimentally investigated using Co-60 irradiation facility and electron beam irradiation facility at Takasaki Radiation Chemistry Research Establishment of JAERI. Nafion is composed of PTFE (Polytetrafluoroethylene) main chain. However the degradation of its mechanical strength by irradiation was significantly distinguished from that of PTFE and no serious damage was observed for its ion exchange capacity up to 530 kGy, the requirement of ITER. Atmospheric effects such as soaking and oxygen on degrading behaviors were discussed from the viewpoint of radical reaction mechanism. Dependencies of operating temperature and radioactive source are also demonstrated in detail.

Journal Articles

Ion and neutron beam analyses of hydrogen isotopes

Kubota, Naoyoshi; Ochiai, Kentaro; Kutsukake, Chuzo; Kondo, Keitaro*; Shu, Wataru; Nishi, Masataka; Nishitani, Takeo

Fusion Engineering and Design, 81(1-7), p.227 - 231, 2006/02

 Times Cited Count:5 Percentile:60.36(Nuclear Science & Technology)

Hydrogen isotopes play important roles in the fuel recycling, the plasma condition etc. at the surface region of plasma facing components. The Fusion Neutronics Source (FNS) of Japan Atomic Energy Research Institute has started microanalysis studies for fusion components since 2002 by applying the beam analyses. In this study, we have measured tritium depth profiles of TFTR tiles exposed to the deuterium-tritium plasma to reveal the hydrogen isotope behavior at the surface region using some microscopic techniques for material analyses at FNS. As the result of the deuteron nuclear reaction analysis, four kinds of elements; deuterium, tritium, lithium-6 and lithium-7, were identified from the energy spectra. Using the spectra, depth profiles of each element were also calculated. The tritium profile had a peak at 0.5 micron, whereas the deuterium and lithium profiles were uniform from the surface to 1.0 micron depth. In addition, the surface region of the TFTR tile has retained the tritium more than one order of magnitude in the bulk.

Journal Articles

Study on tritium accountancy in fusion DEMO plant at JAERI

Nishi, Masataka; Yamanishi, Toshihiko; Hayashi, Takumi; DEMO Plant Design Team

Fusion Engineering and Design, 81(1-7), p.745 - 751, 2006/02

 Times Cited Count:28 Percentile:10.76(Nuclear Science & Technology)

The fusion DEMO plant is under designing at JAERI as a fusion machine following ITER, and it is designed with long-term steady operation and tritium breeding blanket in which more tritium is produced than consumption. Therefore, proper tritium accountancy control concept should be discussed and developed for its safety and operation. From the viewpoint of regulation for the radioisotopes, at first, it will be suitable to divide facilities of the fusion DEMO plant into three accountancy control blocks, that is, (1) the contaminated waste management facility, (2) the long term tritium storage facility, and (3) the fuel processing plant. In each block, tritium amount of receipt and delivery should be carefully accounted. The fuel processing plant involves tritium production in the blanket, therefore proper accounting method for produced tritium should be established. Furthermore, dynamic accountancy is indispensable to the fuel processing plant to monitor tritium inventory distribution for safety and optimum system control in addition to the accountancy under regulation.

Journal Articles

A Design study for tritium recovery system from cooling water of a fusion power plant

Yamanishi, Toshihiko; Iwai, Yasunori; Kawamura, Yoshinori; Nishi, Masataka

Fusion Engineering and Design, 81(1-7), p.797 - 802, 2006/02

no abstracts in English

Journal Articles

Influence of blistering on deuterium retention in tungsten irradiated by high flux deuterium 10-100eV plasmas

Luo, G.; Shu, Wataru; Nishi, Masataka

Fusion Engineering and Design, 81(8-14), p.957 - 962, 2006/02

 Times Cited Count:58 Percentile:2.87(Nuclear Science & Technology)

The influence of blistering on deuterium retention in W was investigated using the newly established plasma generator with controllable incident energies ranging from 100 eV down to around 10 eV and incident flux of 1$$times$$10$$^{22}$$ D/m$$^{2}$$/s. The retention in the irradiated samples was measured using a thermal desorption spectrometer (TDS) at a ramping rate of 5 $$^{circ}$$C/s. The results indicate that only one peak appears in each spectrum, with the peak temperatures ranging from 500 until 850 $$^{circ}$$C, much higher than those from the trapping sites like vacancies, grain boundaries, dislocation loops, or impurities, implying probably a direct origin from the molecules existing inside blisters, voids/bubbles. Significant decrease in the retention at a certain incident fluence after blister appearance was observed and attributed to rupturing of the blisters, consistent with the limited size and increasing number of the blisters with increasing the incident fluence, as observed by means of SEM.

Journal Articles

Monitoring of tritium in diluted gases by detecting bremsstrahlung X-rays

Shu, Wataru; Matsuyama, Masao*; Suzuki, Takumi; Nishi, Masataka

Fusion Engineering and Design, 81(1-7), p.803 - 808, 2006/02

 Times Cited Count:9 Percentile:42.16(Nuclear Science & Technology)

In this work, the counting rate of bremsstrahlung X-rays was measured against the tritium partial pressure in two mixed gases diluted with helium or hydrogen. Subsequently, the counting rate was also measured against total pressure for T$$_{2}$$-He mixture at a constant tritium partial pressure of 93 Pa or 1.3 kPa. For both mixtures, the counting rate of bremsstrahlung X-rays decreased linearly with the decreasing tritium partial pressure when the total pressure is smaller than about 10 kPa. At higher pressures, the deviation from the linear relationship appeared due to absorption of beta-particles in the gas phase, and this can be decreased by some commercially available arrangements. On the other hand, the counting rate of bremsstrahlung X-rays depended only upon the tritium partial pressure when absorption of beta-particles in the gas phase is negligibly small. The results obtained show that this method of tritium monitoring is very promising for the fuel processing system of fusion reactors, especially for tritium recovery system of breeding blankets.

Journal Articles

Characterization of JT-60U exhaust gas during experimental operation

Isobe, Kanetsugu; Nakamura, Hirofumi; Kaminaga, Atsushi; Tsuzuki, Kazuhiro; Higashijima, Satoru; Nishi, Masataka; Kobayashi, Yasunori*; Konishi, Satoshi*

Fusion Engineering and Design, 81(1-7), p.827 - 832, 2006/02

 Times Cited Count:9 Percentile:42.16(Nuclear Science & Technology)

Exhaust gas from JT-60U during experimental operation has been measured with Gas Chromatography (GC), and the gas exhaust characteristic from JT-60U on plasma discharge conditions has been investigated during the JT-60U experimental campaign in 2003-2004. During experimental operation of JT-60U, hydrogen isotope concentration strongly depended on the type of discharges such as high performance, long pulse and so on. On the other hand, impurity species, such as helium, hydrocarbon and carbon oxide, were detected during plasma discharges occasionally. During the experimental operation, plasma disruption remarkably tended to produce high concentration impurities. Glow discharge and Taylor discharge for wall conditioning also produced impurities. In the case of normal plasma, impurity was detected and high performance plasma, such as high $$beta$$ plasma, tended to produce high concentration impurities. This result indicated that impurities concentration might be higher in the case of normal plasma in ITER, because of its high performance.

Journal Articles

Feasibility study on the blanket tritium recovery system using the palladium membrane diffuser

Kawamura, Yoshinori; Enoeda, Mikio; Yamanishi, Toshihiko; Nishi, Masataka

Fusion Engineering and Design, 81(1-7), p.809 - 814, 2006/02

 Times Cited Count:14 Percentile:27.2(Nuclear Science & Technology)

Tritium bred in the solid breeder blanket of a fusion reactor is extracted by passing of a helium sweep gas. Tritium is separated from sweep gas at the blanket tritium recovery system. Palladium membrane diffuser is one of the applicable processes for the blanket tritium recovery system. It is usually applied for hydrogen purification system such as TEP in ITER. However, it has been reported that the rate controlling step changes at lower hydrogen pressure such as the blanket sweep gas condition, and discussion about application for the blanket sweep gas condition is not enough. Recently, conceptual design of the demonstration reactor, named "DEMO2001", has been proposed from JAERI. In this report, the application of the Pd diffuser for the blanket sweep gas condition is discussed based on the condition of DEMO 2001.

Journal Articles

Sorption and desorption of tritiated water on four kinds of materials for ITER

Kobayashi, Kazuhiro; Hayashi, Takumi; Nishi, Masataka; Oya, Yasuhisa*; Okuno, Kenji*

Fusion Engineering and Design, 81(8-14), p.1379 - 1384, 2006/02

 Times Cited Count:5 Percentile:60.36(Nuclear Science & Technology)

In facilities of ITER, various construction materials are possibly exposed by tritium during periodical maintenances and an accident. It is required to establish the effective surface decontamination methods for the above construction materials of ITER. In tritium decontaminating, so-called "soaking" effect is important. This effect is based on sorption of tritiated water on the materials and subsequent desorption from them. In order to obtain and summarize data on the amount of tritium adsorption on the various materials, a series of tritiated water vapor exposure experiments have been carried out as a function of time. The amounts of tritium adsorption on the materials saturated almost within the period from several weeks to 1 month. The adsorption rate of the epoxy was found to be the largest. In the exposure time less than 2 hrs, the adsorption coefficients for the examined materials were found to be in the same order as those reported by F.Ono. It will be also discussed from viewpoint of kinetics for adsorption and desorption.

Journal Articles

Design study of fusion DEMO plant at JAERI

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Sato, Masayasu; Isono, Takaaki; Sakurai, Shinji; Nakamura, Hirofumi; Sato, Satoshi; Suzuki, Satoshi; Ando, Masami; et al.

Fusion Engineering and Design, 81(8-14), p.1151 - 1158, 2006/02

 Times Cited Count:109 Percentile:0.76(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Incident energy dependence of blistering at tungsten irradiated by low energy high flux deuterium plasma beams

Luo, G.; Shu, Wataru; Nishi, Masataka

Journal of Nuclear Materials, 347(1-2), p.111 - 117, 2005/12

 Times Cited Count:69 Percentile:2.27(Materials Science, Multidisciplinary)

Polycrystalline tungsten samples have been irradiated at room temperature by high flux (1$$times$$10$$^{22}$$ D/m$$^{2}$$/s) deuterium plasma beams with incident energies ranging 7 $$sim$$ 98 eV/D. Surface blistering occurred at all energies. The critical fluence for blistering $$Phi$$$$_{cr}$$ was found to increase with decreasing the incident energy. At energies $$<$$ 20 eV/D, $$Phi$$$$_{cr}$$ increased more rapidly. This energy dependence of $$Phi$$$$_{cr}$$ may be explained by a proposed model dealing with the oxide barrier to deuterium uptake into and release from the bulk W. At all energies, the blisters increased in their size and number with fluence within the corresponding low fluence ranges. However the size stopped increasing at certain fluences, while the number kept increasing within the experimental fluence range, which could be attributed to rupturing of blisters at a certain size of about 2$$mu$$m.

JAEA Reports

Design of spiral fin type condenser for hydrogen cryogenic distillation column

Iwai, Yasunori; Yamanishi, Toshihiko; Nishi, Masataka

JAERI-Tech 2005-045, 38 Pages, 2005/08

JAERI-Tech-2005-045.pdf:2.1MB

no abstracts in English

Journal Articles

Tritium recovery from solid breeder blanket by water vapor addition to helium sweep gas

Kawamura, Yoshinori; Iwai, Yasunori; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko; Nishi, Masataka

Fusion Science and Technology, 48(1), p.654 - 657, 2005/07

 Times Cited Count:3 Percentile:73.37(Nuclear Science & Technology)

Adding some amount of hydrogen to the helium sweep gas is effective for tritium extraction from blanket, but it causes permeation of tritium to a cooling system. In the design study of a demonstration reactor in JAERI, tritium leakage has been estimated to be about 20% of bred tritium under typical sweep gas conditions. If these tritiums are recovered under the ITER-WDS condition, tritium leakage limitation has to be less than 0.3% of typical case. Water vapor addition to the sweep gas is effective not only for blanket tritium extraction but also for permeation prevention. The reaction rate of isotope exchange is larger than the case of H$$_2$$, and the equilibrium constant is also expected to be about 1.0. When the H/T ratio is 100, tritium inventory of breeder material is larger than the case of H$$_2$$ addition. However it is not so large. In case of H$$_2$$O sweep, separation of tritiated water from helium seems to be easyer, but the process that changes HTO to HT is necessary.

Journal Articles

Evaluation of tritium behavior in the epoxy painted concrete wall of ITER hot cell

Nakamura, Hirofumi; Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

Fusion Science and Technology, 48(1), p.452 - 455, 2005/07

 Times Cited Count:2 Percentile:80.59(Nuclear Science & Technology)

Tritium behavior released in ITER hot cell has been investigated numerically. Tritium behavior was evaluated by a combined analytical methods of a tritium transport analysis with the one dimensional diffusion model in the multi-layer wall (concrete and epoxy paint) and a tritium concentration analysis with the complete mixing model by the ventilation in the hot cell under the simulated hot cell operational conditions. As the results, tritium concentration in the hot cell volume decreases rapidly from 300 DAC (Derived Air Concentration) less than 1 DAC in several days after removing the tritium release source. Tritium inventory in the wall is estimated to be about 0.1 PBq for 20 years operation. On the other hand, Tritium permeation through the epoxy painted concrete wall will be negligible. Finally, as to the effect of epoxy paint on the tritium permeation and inventory, it is found that the epoxy paint can reduce tritium inventory by about two orders of magnitude relative to bare concrete wall.

Journal Articles

The Oxidation performance test of detritiation system under existence of CO and CO$$_{2}$$

Kobayashi, Kazuhiro; Terada, Osamu*; Miura, Hidenori*; Hayashi, Takumi; Nishi, Masataka

Fusion Science and Technology, 48(1), p.476 - 479, 2005/07

 Times Cited Count:10 Percentile:39.15(Nuclear Science & Technology)

To construct the ITER with high safety and acceptability, it is necessary to establish and to ensure the tritium safe handling technology. The performance of the detritiation system at the off-normal events has not been confirmed well. To obtain performance data of detritiation system at the off normal events, the detritiation experiment was performed at TPL/JAERI using a scaled detritiation system for the oxidation performance test. The detritiation system consists of two oxidation catalyst beds (473K and 773K) and a molecular sieve drying absorber. Basic performance of the detritiation system for hydrogen and methane in air was evaluated under maximum ventilation flow rate. Obtained oxidation efficiency was more than 99.99% for hydrogen in the catalyst bed of 473K and more than 99.9% for methane in the 773K one, respectively. It was confirmed that these performances were maintained even under carbon dioxide , carbon monoxide if oxygen remained in the process gas.

Journal Articles

Radiochemical reactions between tritium molecule and carbon dioxide

Shu, Wataru; Ohira, Shigeru; Suzuki, Takumi; Nishi, Masataka

Fusion Science and Technology, 48(1), p.684 - 687, 2005/07

 Times Cited Count:2 Percentile:80.59(Nuclear Science & Technology)

As part of a series of studies on radiochemical reactions that may take place in the fuel processing systems of a future D-T fusion machine like the ITER, reactions of tritium molecule (T$$_{2}$$) and carbon dioxide (CO$$_{2}$$) were examined by laser Raman spectroscopy and quadrupole mass spectrometry (QMS). Both T$$_{2}$$ and CO$$_{2}$$ decreased rapidly in the first 30 minutes after mixing, and then the reactions between them became much slower. As the predominant products of the reactions, carbon monoxide (CO) and tritiated water (T$$_{2}$$O) were found in gaseous phase and condensed phase, respectively. However, there existed also some solid products that were thermally decomposed to CO, CO$$_{2}$$, T$$_{2}$$, T$$_{2}$$O, etc. during baking at 150$$^{circ}$$C and 250$$^{circ}$$C.

Journal Articles

Tritium release behavior from JT-60U vacuum vessel during air exposure phase and wall conditioning phase

Isobe, Kanetsugu; Nakamura, Hirofumi; Kaminaga, Atsushi; Higashijima, Satoru; Nishi, Masataka; Konishi, Satoshi*; Nishikawa, Masabumi*; Tanabe, Tetsuo*

Fusion Science and Technology, 48(1), p.302 - 305, 2005/07

 Times Cited Count:2 Percentile:80.59(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Interlinked test results for fusion fuel processing and blanket tritium recovery systems using cryogenic molecular sieve bed

Yamanishi, Toshihiko; Hayashi, Takumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Uzawa, Masayuki*; Nishi, Masataka

Fusion Science and Technology, 48(1), p.63 - 66, 2005/07

 Times Cited Count:6 Percentile:56.05(Nuclear Science & Technology)

no abstracts in English

183 (Records 1-20 displayed on this page)