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Journal Articles

Atmospheric modeling of $$^{137}$$Cs plumes from the Fukushima Daiichi Nuclear Power Plant; Evaluation of the model intercomparison data of the Science Council of Japan

Kitayama, Kyo*; Morino, Yu*; Takigawa, Masayuki*; Nakajima, Teruyuki*; Hayami, Hiroshi*; Nagai, Haruyasu; Terada, Hiroaki; Saito, Kazuo*; Shimbori, Toshiki*; Kajino, Mizuo*; et al.

Journal of Geophysical Research; Atmospheres, 123(14), p.7754 - 7770, 2018/07

We compared seven atmospheric transport model results for $$^{137}$$Cs released during the Fukushima Daiichi Nuclear Power Plant accident. All the results had been submitted for a model intercomparison project of the Science Council of Japan in 2014. We assessed model performance by comparing model results with observed hourly atmospheric concentrations of $$^{137}$$Cs, focusing on nine plumes over the Tohoku and Kanto regions. The results showed that model performance for $$^{137}$$Cs concentrations was highly variable among models and plumes. We also assessed model performance for accumulated $$^{137}$$Cs deposition. Simulated areas of high deposition were consistent with the plume pathways, though the models that best simulated $$^{137}$$Cs concentrations were different from those that best simulated deposition. The ensemble mean of all models consistently reproduced $$^{137}$$Cs concentrations and deposition well, suggesting that use of a multimodel ensemble results in more effective and consistent model performance.

Journal Articles

Fuel behavior analysis for accident tolerant fuel with sic cladding using adapted FEMAXI-7 code

Shirasu, Noriko; Saito, Hiroaki; Yamashita, Shinichiro; Nagase, Fumihisa

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 8 Pages, 2017/09

Silicon carbide (SiC) is an attractive candidate of accident tolerant fuel (ATF) cladding material because of its high chemical stability, high radiation resistance and low neutron absorption. FEMAXI-ATF has been developed to analysis SiC cladding fuel behaviors. The thermal, mechanical and irradiation property models were implemented to FEMAXI-7, which is a fuel behavior analysis code being developed in JAEA. Fuel rod behavior analysis was performed under typical boiling water reactor (BWR) operating conditions with a model based on a 9$$times$$9 BWR fuel (Step III Type B), in which the cladding material was replaced from Zircaloy to SiC. The SiC cladding shows large swelling by irradiation. It increases the gap size and decreases cladding thermal conductivity. The mechanism of relaxation of stress is also different from the Zircaloy cladding. The experimental data for SiC materials are still insufficient to construct the models, especially for evaluating fracture behavior.

Journal Articles

The Applicability of SiC-SiC fuel cladding to conventional PWR power plant

Furumoto, Kenichiro*; Watanabe, Seiichi*; Yamamoto, Teruhisa*; Teshima, Hideyuki*; Yamashita, Shinichiro; Saito, Hiroaki; Shirasu, Noriko

Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09

Since 2015, Mitsubishi Nuclear Fuel (MNF) has joined in a Japanese R&D project of ATF founded by the Ministry of Economy, Trade and Industry (METI) as a subcontractor to Japan Atomic Energy Agency (JAEA) which is the prime contractor to METI. In this program, MNF plans to evaluate an influence of Silicon Carbide (SiC) composite cladding upon fuel rod behavior in current pressurized water reactors (PWR). This paper reports the evaluation result of the applicability of fuel rod with SiC composite cladding for a conventional PWR. For the applicability evaluations of SiC composite to conventional PWR, both of analytical evaluations and out-of-pile tests for SiC composite were conducted. Analytical evaluations were performed by Mitsubishi's own fuel rod design code and the fuel rod behavior evaluation code developed by JAEA. These codes were modified to evaluate the behavior of the fuel rod with SiC composite cladding. As out-of-pile tests, thermal diffusivity measurement and autoclave corrosion test for SiC composite samples were performed. Test apparatus were developed for evaluation of performance of SiC composite under the condition simulated design basis accident (DBA).

Journal Articles

Study on the deterioration mechanism of layered rock-salt electrodes using epitaxial thin films; Li(Ni, Co, Mn)O$$_{2}$$ and their Zr-O surface modified electrodes

Abe, Machiko*; Iba, Hideki*; Suzuki, Kota*; Minamishima, Hiroaki*; Hirayama, Masaaki*; Tamura, Kazuhisa; Mizuki, Junichiro*; Saito, Tomohiro*; Ikuhara, Yuichi*; Kanno, Ryoji*

Journal of Power Sources, 345, p.108 - 119, 2017/03

 Percentile:100(Chemistry, Physical)

The surface structure of the Li(Ni, Co, Mn)O$$_{2}$$ electrode was studied during charge/discharge process using electrochemical methods and X-ray/Neutron scattering techniques. It was found that during charge/discharge process the coverage of spinel structure increased. The spinel structure has low electrochemical activity and is not involved in Li insertion/extraction. After the surface modification, it was found that the coverage of the spinel structure did not increase. Further, it was also found out that the Li concentration at the electrode/electrolyte interface increased.

Journal Articles

Utilization of $$^{134}$$Cs/$$^{137}$$Cs in the environment to identify the reactor units that caused atmospheric releases during the Fukushima Daiichi accident

Chino, Masamichi; Terada, Hiroaki; Nagai, Haruyasu; Katata, Genki; Mikami, Satoshi; Torii, Tatsuo; Saito, Kimiaki; Nishizawa, Yukiyasu

Scientific Reports (Internet), 6, p.31376_1 - 31376_14, 2016/08

 Times Cited Count:10 Percentile:8.68(Multidisciplinary Sciences)

Journal Articles

Lattice structure transformation and change in surface hardness of Ni$$_3$$Nb and Ni$$_3$$Ta intermetallic compounds induced by energetic ion beam irradiation

Kojima, Hiroshi*; Yoshizaki, Hiroaki*; Kaneno, Yasuyuki*; Semboshi, Satoshi*; Hori, Fuminobu*; Saito, Yuichi; Okamoto, Yoshihiro; Iwase, Akihiro*

Nuclear Instruments and Methods in Physics Research B, 372, p.72 - 77, 2016/04

 Times Cited Count:3 Percentile:42.29(Instruments & Instrumentation)

Ni$$_3$$Nb and Ni$$_3$$Ta intermetallic compounds, which show the complicated lattice structures were irradiated with 16 MeV Au$$^{5+}$$ ions at room temperature. The X-ray diffraction measurement revealed that the lattice structure of these intermetallic compounds changed from the ordered structures to the amorphous state by the ion irradiation. The irradiation-induced amorphization caused the increase in Vickers hardness. The result was compared with our previous results for Ni$$_3$$Al and Ni$$_3$$V, and was discussed in terms of the intrinsic lattice structures of the samples.

Journal Articles

Matrix diffusion and sorption of Cs$$^{+}$$, Na$$^{+}$$, I$$^{-}$$ and HTO in granodiorite; Laboratory-scale results and their extrapolation to the in situ condition

Tachi, Yukio; Ebina, Takanori*; Takeda, Chizuko*; Saito, Toshihiko*; Takahashi, Hiroaki*; Ouchi, Yuji*; Martin, A. J.*

Journal of Contaminant Hydrology, 179, p.10 - 24, 2015/08

 Times Cited Count:16 Percentile:15.73(Environmental Sciences)

Matrix diffusion and sorption are important processes in the assessment of radionuclide transport in crystalline rocks. Diffusion and sorption parameters for Cs$$^{+}$$, Na$$^{+}$$, I$$^{-}$$ and HTO were determined by through-diffusion and batch sorption experiments using granodiorite samples from the Grimsel Test Site, Switzerland. The De values were in the order Cs$$^{+}$$, Na$$^{+}$$, HTO, I$$^{-}$$. The capacity factor and Kd values show the same trends. The dual depth profiles for Cs$$^{+}$$ and Na$$^{+}$$ can be interpreted by a near-surface Kd increment. The microscopic analysis indicated that this is caused by high porosity and sorption capacities in disturbed biotite minerals on the sample surface. The Kd values derived from the dual profiles are likely to correspond to Kd dependence on the grain sizes of crushed samples in the batch experiments. The results of the in situ LTD experiments were interpreted reasonably well by using transport parameters derived from laboratory data and extrapolating them to in situ conditions.

Journal Articles

Source term estimation for the Fukushima Daiichi Nuclear Power Station accident by combined analysis of environmental monitoring and plant data through atmospheric dispersion simulation

Nagai, Haruyasu; Terada, Hiroaki; Chino, Masamichi; Katata, Genki; Mikami, Satoshi; Saito, Kimiaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4044 - 4052, 2015/08

JAEA has estimated the atmospheric releases of radionuclide during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric transport and deposition model (ATDM). To improve our source term, we are trying to develop more sophisticated estimation method and use new information from severe accident analysis and observation data. As the first step of new trial, we used $$^{134}$$Cs/$$^{137}$$Cs ratios of inventories in FNPS1 reactors Unit 1 to 3 and those in surface deposition. By considering temporal change in $$^{134}$$Cs/$$^{137}$$Cs ratio of released plume and ATDM simulations, spatial distribution of $$^{134}$$Cs/$$^{137}$$Cs ratio in surface deposition was explained. This result can be used to specify from which reactor the dominant release occurred for each time period, and consequently provide useful information to severe accident analysis for the FNPS1 case.

Journal Articles

Spatial distributions of radionuclides deposited onto ground soil around the Fukushima Dai-ichi Nuclear Power Plant and their temporal change until December 2012

Mikami, Satoshi; Maeyama, Takeshi*; Hoshide, Yoshifumi*; Sakamoto, Ryuichi*; Sato, Shoji*; Okuda, Naotoshi*; Demongeot, S.*; Gurriaran, R.*; Uwamino, Yoshitomo*; Kato, Hiroaki*; et al.

Journal of Environmental Radioactivity, 139, p.320 - 343, 2015/01

 Times Cited Count:36 Percentile:6.33(Environmental Sciences)

Journal Articles

Development of operation and maintenance technology for HTGRs by using HTTR (High Temperature engineering Test Reactor)

Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Homma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Takada, Shoji; Shinozaki, Masayuki

Nuclear Engineering and Design, 271, p.499 - 504, 2014/05

 Times Cited Count:3 Percentile:61.51(Nuclear Science & Technology)

The long term high temperature operation using HTTR was carried out to establish the technical basis of HTGR in the high temperature test operation mode during 50-day since January till March, 2010. It is necessary to demonstrate the stability of plant during long-term operation in order to attain the stable supply of the high temperature heat to the planned heat utilization system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, Stability and reliability of the components and facility was demonstrated by evaluating the heat transfer performance of high temperature components, the performance of pressure control to compensate helium gas leak, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the operation, the technical basis for the operation and maintenance technology of HTGRs was established.

JAEA Reports

Light water reactor fuel analysis code FEMAXI-7; Model and structure (Revised edition)

Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Amaya, Masaki

JAEA-Data/Code 2013-014, 382 Pages, 2014/03

JAEA-Data-Code-2013-014.pdf:16.36MB

A light water reactor fuel analysis code FEMAXI-7 has been developed as the latest version for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. This report is the revised edition of the first one, JAEA-Data/Code 2010-035, which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition was extended by orderly addition and disposition of explanations of models and organized as this revised edition after three years interval.

Journal Articles

The Influence on the vulcanized rubber physical properties by radiation grafting

Saito, Hiroaki*; Mizote, Norihito*; Ueki, Yuji; Seko, Noriaki

JAEA-Review 2013-059, JAEA Takasaki Annual Report 2012, P. 57, 2014/03

no abstracts in English

JAEA Reports

Development of safety management system for works in radiation controlled area (Joint research)

Hiyama, Kazuhisa; Hanawa, Nobuhiro; Kurosawa, Akihiko; Eguchi, Shohei; Hori, Naohiko; Kusunoki, Tsuyoshi; Ueda, Hisao; Shimada, Hiroshi; Kanda, Hiroaki*; Saito, Isamu*

JAEA-Technology 2013-045, 32 Pages, 2014/02

JAEA-Technology-2013-045.pdf:5.83MB

This report summarizes regarding to develop of real-time multifunctional access control system which is able to manage worker's access control and exposure dose at real-time in the reactor building, besides worker's location and worker might be fall down by accident.

JAEA Reports

Research on engineering technology in the full-scale demonstration of EBS and operation technology for HLW disposal; Research report in 2011 (Joint research)

Nakatsuka, Noboru; Sato, Haruo; Tanai, Kenji; Sugita, Yutaka; Nakayama, Masashi; Sawada, Sumiyuki*; Niinuma, Hiroaki*; Asano, Hidekazu*; Saito, Masahiko*; Yoshino, Osamu*; et al.

JAEA-Research 2013-027, 34 Pages, 2013/11

JAEA-Research-2013-027.pdf:5.84MB

Japan Atomic Energy Agency (JAEA) and Radioactive Waste Management Funding and Research Center (RWMC) concluded the letter of cooperation agreement on the research and development of radioactive waste disposal in April, 2005, and have been carrying out the collaboration work based on the agreement. JAEA have been carrying out the Horonobe Underground Research Laboratory (URL) Project which is intended for a sedimentary rock in the Horonobe town, Hokkaido, since 2001. In the project, geoscientific research and research and development on geological disposal technology are being promoted. Meanwhile, the government (the Agency for Natural Resources and Energy, Ministry of Economy, Trade and Industry) has been promoting construction of equipments for the full-scale demonstration of engineered barrier system and operation technology for high-level radioactive waste (HLW) disposal since 2008, to enhance public's understanding to the geological disposal of HLW, e.g. using underground facility. RWMC received an order of the project in fiscal year 2010 (2010/2011) continuing since fiscal year 2008 (2008/2009). Since topics in this project are included in the Horonobe URL Project, JAEA carried out this project as collaboration work continuing in fiscal year 2008. This report summarizes the results of engineering technology carried out in this collaboration work in fiscal year 2011. In fiscal year 2011, part of the equipments for emplacement of buffer material was produced and visualization test for water penetration in buffer material were carried out.

JAEA Reports

Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa

JAEA-Data/Code 2013-009, 306 Pages, 2013/10

JAEA-Data-Code-2013-009.pdf:5.73MB

A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7.

Journal Articles

Report of the Special Symposium on the Transport and Diffusion of Contaminants from the Fukushima Dai-ichi Nuclear Power Plant; Present status and future directions

Kondo, Hiroaki*; Yamada, Tetsuji*; Chino, Masamichi; Iwasaki, Toshiki*; Katata, Genki; Maki, Takashi*; Saito, Kazuo*; Terada, Hiroaki; Tsuruta, Haruo*

Tenki, 60(9), p.723 - 729, 2013/09

AA2013-0745.pdf:0.51MB

no abstracts in English

JAEA Reports

Light water reactor fuel analysis code FEMAXI-7; Model and structure

Suzuki, Motoe; Saito, Hiroaki*; Udagawa, Yutaka; Nagase, Fumihisa

JAEA-Data/Code 2013-005, 382 Pages, 2013/07

JAEA-Data-Code-2013-005.pdf:6.4MB

A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions.

Journal Articles

Development of small specimen test techniques for the IFMIF test cell

Wakai, Eiichi; Kim, B. J.; Nozawa, Takashi; Kikuchi, Takayuki; Hirano, Michiko*; Kimura, Akihiko*; Kasada, Ryuta*; Yokomine, Takehiko*; Yoshida, Takahide*; Nogami, Shuhei*; et al.

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 6 Pages, 2013/03

Journal Articles

Surface modification of vulcanized rubber by radiation grafting

Saito, Hiroaki*; Mizote, Norihito*; Ueki, Yuji; Seko, Noriaki

JAEA-Review 2012-046, JAEA Takasaki Annual Report 2011, P. 45, 2013/01

no abstracts in English

Journal Articles

Development of operation and maintenance technology of HTTR (High Temperature engineering Test Reactor)

Shimizu, Atsushi; Kawamoto, Taiki; Tochio, Daisuke; Saito, Kenji; Sawahata, Hiroaki; Homma, Fumitaka; Furusawa, Takayuki; Saikusa, Akio; Shinozaki, Masayuki

Proceedings of 6th International Topical Meeting on High Temperature Reactor Technology (HTR 2012) (USB Flash Drive), 8 Pages, 2012/10

To establish the technical basis of HTGR, the long term high temperature operation using HTTR was carried out during 50-day in 2010. It is necessary to demonstrate the stability of plant during long-term operation and the reliability of components and facilities special to HTGRs, in order to attain the stable supply of the high temperature heat to the planned hydrogen production system of HTTR. Test data obtained in the operation were evaluated for the technical issues which were extracted before the operation. As the results, stability and reliability of the components and facility special to HTGRs was demonstrated by evaluating the heat transfer performance of high temperature components, the helium gas leak tightness, the reliability of the dynamic components such as helium gas circulators, the performance of heat-up protection of radiation shielding. Through the operation, the technical basis for the operation and maintenance technology of HTGRs were established.

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