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Journal Articles

Plutonium dioxide particle imaging using a high-resolution alpha imager for radiation protection

Morishita, Yuki; Kurosawa, Shunsuke*; Yamaji, Akihiro*; Hayashi, Masateru*; Sasano, Makoto*; Makita, Taisuke*; Azuma, Tetsushi*

Scientific Reports (Internet), 11(1), p.5948_1 - 5948_11, 2021/03

 Times Cited Count:0 Percentile:100(Multidisciplinary Sciences)

The internal exposure of workers who inhale plutonium dioxide particles in nuclear facilities is a crucial matter for human protection from radiation. To determine the activity median aerodynamic diameter values at the working sites of nuclear facilities in real time, we developed a high-resolution alpha imager using a ZnS(Ag) scintillator sheet, an optical microscope, and an electron-multiplying charge-coupled device camera. Then, we designed and applied a setup to measure a plutonium dioxide particle and identify the locations of the individual alpha particles in real time. Employing a Gaussian fitting, we evaluated the average spatial resolution of the multiple alpha particles was evaluated to be 16.2 umFWHM with a zoom range of 5 x. Also, the spatial resolution for the plutonium dioxide particle was 302.7 umFWHM due to the distance between the plutonium dioxide particle and the ZnS(Ag) scintillator. The influence of beta particles was negligible, and alpha particles were discernible in the alpha-beta particle contamination. The equivalent volume diameter of the plutonium dioxide particle was calculated from the measured count rate. These results indicate that the developed alpha imager is effective in the plutonium dioxide particle measurements at the working sites of nuclear facilities for internal exposure dose evaluation.

JAEA Reports

Design study for impermeable function of trench disposal facility for very low level waste generated from research, industrial and medical facilities (Joint research)

Sakai, Akihiro; Kurosawa, Ryohei*; Nakata, Hisakazu; Okada, Shota; Izumo, Sari; Sato, Makoto*; Kitamura, Yoichi*; Honda, Yasutake*; Takaoka, Katsuki*; Amazawa, Hiroya

JAEA-Technology 2016-019, 134 Pages, 2016/10

JAEA-Technology-2016-019.pdf:8.25MB

Japan Atomic Energy Agency has been developing to design trench disposal facility with impermeable layers in order to dispose of miscellaneous waste. Geomembrane liners have a function that prevent seepage of leachant and collect the leachant. However, the geomembrane liners do not necessarily provide the expected performance due to damage generated when heavy equipment contacts with the liner. Therefore, we studied the impermeable layers having high performance of preventing seepage of leachant including radioactivity taking into account characteristics of low permeable materials and effect of multiple layer structure. As results, we have evaluated that the composite layers composed by a drainage layer, geomembrane liners and a low permeable layer are most effective structure to prevent seepage of leachant. Taking into account disposal of waste including cesium, we also considered zeolite containing sheets for adsorption of cesium were installed in the impermeable layers.

Journal Articles

Demonstrative experiments on the migration of radiocesium from buried soil contaminated by the accident at Fukushima Daiichi Nuclear Power Station

Yamaguchi, Tetsuji; Shimada, Taro; Ishibashi, Makoto*; Akagi, Yosuke*; Kurosawa, Mitsuru*; Matsubara, Akiyoshi*; Matsuda, Yuki*; Sato, Shigeyoshi*

Genshiryoku Bakkuendo Kenkyu (CD-ROM), 22(2), p.21 - 27, 2015/12

It is predictable from previous studies that radiocesium hardly migrate into surrounding soils and groundwater from soils contaminated by the Fukushima Daiichi Nuclear Power Plant accident if they are buried and covered with indigenous soils. This study demonstrated the prediction by performing in-situ migration experiments over a year in a public park in Miho, Ibaraki prefecture and in two public parks in Misato, Saitama prefecture. Contaminated soils were buried at a depth range of 0.3 - 1.0 m or at 0.3 - 1.3 m and covered with indigenous soil layer of 0.3 m, and were sprinkled with water to accelerate the radiocesium migration. Migration of radiocesium was not observed from radiometric analyses of boring cores and soil water samples. Laboratory column and sorption experiments revealed that the radiocesium hardly leach out of the soil and even if they leach out from the contaminated soil, radiocesium is sorbed on surrounding soils and hardly migrate through the soli layer. Simulation of Cs-137 migration for 100 years by an advection-diffusion model showed that Cs-137 hardly migrate and decay out in the contaminated soil.

Journal Articles

Neutron irradiation effect of high-density MoO$$_{3}$$ pellets for Mo-99 production, 2

Nishikata, Kaori; Ishida, Takuya; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Kimura, Akihiro; Matsui, Yoshinori; Tsuchiya, Kunihiko; Sano, Tadafumi*; Fujihara, Yasuyuki*; et al.

KURRI Progress Report 2014, P. 109, 2015/07

As one of effective applications of the Japan Materials Testing Reactor (JMTR), JAEA has a plan to produce $$^{99}$$Mo by (n,$$gamma$$) method ((n,$$gamma$$)$$^{99}$$Mo production), a parent nuclide of $$^{99m}$$Tc. In this study, preliminary irradiation test was carried out with the high-density molybdenum trioxide (MoO$$_{3}$$) pellets in the hydraulic conveyer (HYD) of the Kyoto University Research Reactor (KUR) and the $$^{99m}$$Tc solution extracted from $$^{99}$$Mo was evaluated. After the irradiation test of the high-density MoO$$_{3}$$ pellets in the KUR, $$^{99m}$$Tc was extracted from the Mo solution and the recovery rate of $$^{99m}$$Tc achieved the target values. The $$^{99m}$$Tc solution also got the value that satisfied the standard value for $$^{99m}$$Tc radiopharmaceutical products by the solvent extraction method.

Journal Articles

Neutron irradiation effect of high-density MoO$$_{3}$$ pellets for Mo-99 production

Nishikata, Kaori; Ishida, Takuya; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Kimura, Akihiro; Matsui, Yoshinori; Tsuchiya, Kunihiko; Sano, Tadafumi*; Fujihara, Yasuyuki*; et al.

KURRI Progress Report 2013, P. 242, 2014/10

As one of effective applications of the Japan Materials Testing Reactor (JMTR), JAEA has a plan to produce Mo-99 ($$^{99}$$Mo) by (n,$$gamma$$) method ((n,$$gamma$$)$$^{99}$$Mo production), a parent nuclide of $$^{99m}$$Tc. In this study, preliminary irradiation tests were carried out with the high-density MoO$$_{3}$$ pellets in the KUR and the $$^{99}$$Mo production amount was evaluated between the calculation results and measurement results.

Journal Articles

Renewal of monitoring boards in control room at the hot laboratory

Kurosawa, Makoto; Kato, Yoshiaki; Yonekawa, Minoru; Taguchi, Taketoshi

UTNL-R-0486, p.9_1 - 9_11, 2014/03

It has been irradiated in the concrete cell, the microscope lead cell, the lead cell for materials examinations and the iron cell and, in the JMTR hot laboratory facilities, examines it after the irradiation such as fuel and nuclear reactor structure materials. I install a monitoring board for a concrete cell, a microscope lead cell, a lead cell for materials examinations and iron cells in the control room I watch concentration such as the minus number pressure in these each cell, the air absorption dose rate in the cell, the cover door opening and shutting indication and to control it. As for these monitoring boards, about 30 through 40 or more passed after an in-service start, and high aging decided to update it in consideration of the driving of approximately 20 years after JMTR re-operation because trouble by becoming it and outbreak of the malfunction were concerned about.

Journal Articles

Mo recycling property from generator materials with irradiated molybdenum

Kakei, Sadanori*; Kimura, Akihiro; Niizeki, Tomotake*; Ishida, Takuya; Nishikata, Kaori; Kurosawa, Makoto; Yoshinaga, Hideo*; Hasegawa, Yoshio*; Tsuchiya, Kunihiko

Proceedings of 6th International Symposium on Material Testing Reactors (ISMTR-6) (Internet), 7 Pages, 2013/10

The Japan Materials Testing Reactor (JMTR) is expected to contribute to the expansion of industrial utilization, such as the domestic production of $$^{99}$$Mo for the medical diagnosis medicine $$^{rm 99m}$$Tc. Production by the (n, $$gamma$$) method is proposed as domestic $$^{99}$$Mo production in JMTR because of the low amount of radioactive wastes and the easy $$^{99}$$Mo/$$^{rm 99m}$$Tc production process. Molybdenum oxide (MoO$$_{3}$$) pellets, poly zirconium compounds (PZC) and poly titanium compounds (PTC) are used as the irradiation target and generator for the production of $$^{99}$$Mo/$$^{rm 99m}$$Tc by the (n, $$gamma$$) method. However, it is necessary to use the enriched $$^{98}$$MoO$$_{3}$$, which is very expensive, to increase the specific activity of $$^{99}$$Mo. Additionally, a large amount of used PZC and PTC is generated after the decay of $$^{99}$$Mo. Therefore, this recycling technology of used PZC/PTC has been developed to recover molybdenum (Mo) as an effective use of resources and a reduction of radioactive wastes. The total Mo recovery rate of this process was 95.8%. From the results of the hot experiments, we could demonstrate that the recovery of MoO$$_{3}$$ and the recycling of PZC are possible. In the future, the equipment of recovering Mo will be installed in JMTR-Hot Cell, and this recycling process will be able to contribute to the reduction of production costs of $$^{rm 99m}$$Tc and the reduction of radioactive wastes.

Journal Articles

Development of post-irradiation test facility for domestic production of $$^{99}$$Mo

Taguchi, Taketoshi; Yonekawa, Minoru; Kato, Yoshiaki; Kurosawa, Makoto; Nishikata, Kaori; Ishida, Takuya; Kawamata, Kazuo

UTNL-R-0483, p.10_5_1 - 10_5_13, 2013/03

JMTR focus on the activation method. By carrying out the preliminary tests using irradiation facilities existing, and verification tests using the irradiation facility that has developed in the cutting-edge research and development strategic strengthening business, as irradiation tests towards the production of $$^{99}$$Mo, we have been conducting research and development that can contribute to supply about 25% for $$^{99}$$Mo demand in Japan and the stable supply of radiopharmaceutical. This report describes a summary of the status of the preliminary tests for the production of $$^{99}$$Mo: Maintenance of test equipment in the facility in JMTR hot laboratory in preparation for research and development for the production of $$^{99}$$Mo in JMTR and using MoO$$_{3}$$ pellet irradiated at Kyoto University Research Reactor Institute (KUR).

JAEA Reports

Study on evaluation method of potential impact of natural phenomena on a HLW disposal system

Kawamura, Makoto; Oi, Takao; Niizato, Tadafumi; Yasue, Kenichi; Tokiwa, Tetsuya; Niwa, Masakazu; Shimada, Koji; Kurosawa, Hideki; Asamori, Koichi; Kawachi, Susumu; et al.

JAEA-Research 2008-018, 47 Pages, 2008/03

JAEA-Research-2008-018.pdf:24.18MB

In this report, we sophisticated the framework as a part of the total system performance assessment for two purposes: the first one is quantification of relationship of characteristic of natural phenomena between geological environmental conditions (THMCG), and the other one is quantification of relationship of THMCG condition between parameters of performance assessment. On the other hand, we applied the sophisticated framework to all natural phenomena. As a result, to apply the sophisticated framework, we could show that information integration could carry out efficiently. Moreover, we have checked that the framework was applicable to all phenomena. Furthermore, we could show that suitable scenarios might be chosen by information integration.

Journal Articles

REIDAC; A Software package for retrospective dose assessment in internal contamination of radionuclides

Kurihara, Osamu; Hato, Shinji; Kanai, Katsuta; Takada, Chie; Takasaki, Koji; Ito, Kimio; Ikeda, Hiroshi*; Oeda, Mikihiro*; Kurosawa, Naohiro*; Fukutsu, Kumiko*; et al.

Journal of Nuclear Science and Technology, 44(10), p.1337 - 1346, 2007/10

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

In the case of internal contamination of radionuclides, it is necessary to perform internal dose assessment for radiation protection. For this purpose, the ICRP has given the dose coefficients and the retention and excretion rates for various radionuclides. However, these dosimetric quantities are calculated only in typical conditions, therefore, are not necessarily covered enough in the case of dose assessment in which specific information on the incident or/and individual biokinetic characteristics should be taken into account retrospectively. This paper describes a developed PC-based package of software REIDAC to meet the needs in retrospective dose assessment. REIDAC was verified by comparisons with dosimetric quantities given on the ICRP publications and several examples of practical use were also shown.

JAEA Reports

Study of sub-surface disposal concepts for uranium waste, 3

Tsujimura, Seiichi; Funabashi, Hideyuki; Ishibashi, Makoto*; Takase, Toshio*; Kurosawa, Mitsuru*

JAEA-Research 2007-030, 105 Pages, 2007/03

JAEA-Research-2007-030.pdf:7.72MB

Uranium waste has characteristics that it is rarely expected to decay its radioactivity and it is not almost recessary to consider external exposure to radiation from waste package. We studied resonable sub-surface disposal concepts for uranium waste considering the characteristics. It showed feasibility of this system. In 2006, the study was done to evaluate the correlation between dose to the public and the parameter in consideration of the uncertainty of the parameter by comprehensive sensitivity analysis calculating repeatedly dose with the data sets by random sampling in parameter ranges given adequately, because last year's study was an evaluation intended for a limited site. The result was that two parameters, "flow velocity of underground water of repository neighborhood" and "distribution coefficient of uranium in natural barrier", had correlation with dose to the public.

JAEA Reports

Study of sub-surface disposal concepts for uranium waste, 2

Tsujimura, Seiichi; Funabashi, Hideyuki; Ishibashi, Makoto*; Takase, Toshio*; Kurosawa, Mitsuru*

JAEA-Research 2006-029, 96 Pages, 2006/07

JAEA-Research-2006-029.pdf:3.97MB

Uranium waste has characteristics that it is rarely expected to decay its radioactivities and it is not almost necessary to consider external exposure to radiation from waste package. We studied reasonable sub-surface disposal concepts for uranium waste in 2004 and 2005 considering the characteristics. In 2005, we studied necessity of engineered barrier for the disposal of uranium waste, considering change of chemical condition around disposal facilities over long periods of time. Safety assessment was made to analyze effect of difference in sorption parameters at reduction and oxidation conditions. The assessment showed that change from reduction to oxidation around disposal facilities did not lead to increase dose to the public. The assessment with realistic sorption parameters showed that dose to the public was not more than 10 $$mu$$Sv/y. The results proved that it was not necessary to keep reduction conditions around disposal facilities. This two-year- study showed that there was possibility of sub-surface disposal system without engineered barrier for uranium waste.

JAEA Reports

Design and installation of high-temperature ultrasonic measuring system and grinder for nuclear fuel containing trans-uranium elements

Serizawa, Hiroyuki; Kikuchi, Hironobu; Iwai, Takashi; Arai, Yasuo; Kurosawa, Makoto; Mimura, Hideaki; Abe, Jiro

JAERI-Tech 2005-039, 23 Pages, 2005/07

JAERI-Tech-2005-039.pdf:2.89MB

A high-temperature ultrasonic measuring system had been designed and installed in a glovebox (711-DGB) to study a mechanical property of nuclear fuel containing trans-uranium (TRU) elements. A figuration apparatus for the cylinder-type sample preparation had also been modified and installed in an established glovebox (142-D). The system consists of an ultrasonic probe, a heating furnace, cooling water-circulating system, a cooling air compressor, vacuum system, gas supplying system and control system. An A/D converter board and an pulsar/Receiver board for the measurement of wave velocity were installed in a personal computer. The apparatus was modified to install into the glovebox. Some safety functions were supplied to the control system. The shape and size of the sample was revised to minimize the amount of TRU elements for the use of the measurement. The maximum sample temperature is 1500 $$^{circ}$$C. The performance of the installed apparatuses and the glovebox were confirmed through a series of tests.

JAEA Reports

Development of nondestructive measurement technique in $$alpha$$-waste including neptunium

Kurosawa, Makoto; Ouchi, Shoichi*; Abe, Jiro; Okane, Shogo; Usui, Takeshi

JAERI-Tech 2002-036, 24 Pages, 2002/03

JAERI-Tech-2002-036.pdf:1.0MB

Method for measurement of plutonium in $$alpha$$-waste using passive $$gamma$$ ray has been adopted at Plutonium Fuel Research Facility. Recently, research of neptunium has been started. It is necessary to evaluate the passive $$gamma$$ ray method for measurement of plutonium in $$alpha$$-waste including plutonium and neptunium together. This report describes the results of comparing the two methods, deduction method and division method. The error of deduction method is about 10 to 15% when the quantity of plutonium is larger than 100 mg in the waste. But the error reaches more than 50% when the quantity of plutonium is less than 10 mg and the ratio of plutonium and neptunium is less than one. On the other hand, the error of division method is about a few to 15% when the quantity of plutonium is larger than 100 mg in the waste. The error is about 30 to 50% when the quantity of plutonium is less than 10 mg, but the error is not affected by the ratio of plutonium and neptunium. By reason of this results, the division method is adopted for measurement of plutonium in $$alpha$$-waste including neptunium.

JAEA Reports

Scrap of gloveboxes No.801-W and No.802-W

Ouchi, Shoichi*; Kurosawa, Makoto; Abe, Jiro; Okane, Shogo; Usui, Takeshi

JAERI-Tech 2002-026, 35 Pages, 2002/03

JAERI-Tech-2002-026.pdf:2.32MB

Both gloveboxes No.801-W for measuring samples of uranium or plutonium and No.802-W for analyzing the quantity of uranium or plutonium are established at twenty five years ago in the analyzing room No.108 of Plutonium Fuel Research Facility. It was planned to scrap the gloveboxes and to establish new gloveboxes. This report describes the technical view of the scrapping works.

JAEA Reports

Seismic analysis of plutonium glovebox by MSC/NASTRAN

Hirata, Masaru; Ishikawa, Kazuya*; Kurosawa, Makoto; ; Hoshina, Hirofumi*

JAERI-M 92-206, 50 Pages, 1993/01

JAERI-M-92-206.pdf:1.62MB

no abstracts in English

Journal Articles

Preparation of pure tritium for a liquid D$$_{2}$$/T$$_{2}$$ target of muon-catalyzed fusion experiments

Kudo, Hiroshi; Fujie, Makoto; Tanase, Masakazu; Kato, Mineo; Kurosawa, Kiyoyuki; Sugai, Hiroyuki; Umezawa, Hirokazu; Matsuzaki, Teiichiro*; Nagamine, Kanetada*

Applied Radiation and Isotopes, 43(5), p.577 - 583, 1992/00

no abstracts in English

Journal Articles

X-ray observation of $$alpha$$-sticking phenomena in muon catalyzed fusion for a high density D-T mixture with 30% tritium concentration

Nagamine, Kanetada*; Matsuzaki, Teiichiro*; Ishida, Katsuhiko*; *; *; *; Miyake, Yasuhiro*; *; *; *; et al.

Muon Catal. Fusion, 5-6, p.289 - 295, 1991/00

no abstracts in English

Journal Articles

Radiogas-chromatographic determination of chemical and isotopic purity of tritium gas in tritium production

Tanase, Masakazu; Kurosawa, Kiyoyuki; Fujie, Makoto; Sugai, Hiroyuki; Okane, Shogo; Kato, Mineo

Fusion Technology, 14, p.1090 - 1095, 1988/09

no abstracts in English

Journal Articles

Production of 40 TBq tritium using neutron-irradiated $$^{6}$$Li-Al alloy

; ; Kurosawa, Kiyoyuki; Motoishi, Shoji; ; ; Fujie, Makoto; ;

Journal of Nuclear Science and Technology, 25(2), p.198 - 203, 1988/02

no abstracts in English

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