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JAEA Reports

Annual report for research on geosphere stability for long-term isolation of radioactive waste in fiscal year 2017

Ishimaru, Tsuneari; Ogata, Nobuhisa; Shimada, Akiomi; Asamori, Koichi; Kokubu, Yoko; Niwa, Masakazu; Watanabe, Takahiro; Saiga, Atsushi; Sueoka, Shigeru; Komatsu, Tetsuya; et al.

JAEA-Research 2018-015, 89 Pages, 2019/03

JAEA-Research-2018-015.pdf:14.43MB

This annual report documents the progress of research and development (R&D) in the 3rd fiscal year during the JAEA 3rd Mid- and Long-term Plan (fiscal years 2015-2021) to provide the scientific base for assessing geosphere stability for long-term isolation of the high-level radioactive waste. The planned framework is structured into the following categories: (1) Development and systematization of investigation techniques, (2) Development of models for long-term estimation and effective assessment, (3) Development of dating techniques. In this report, the current status of R&D activities with previous scientific and technological progress is summarized.

JAEA Reports

Research plan on geosphere stability for long-term isolation of radioactive waste; Scientific Program for fiscal year 2018

Ishimaru, Tsuneari; Ogata, Nobuhisa; Shimada, Akiomi; Asamori, Koichi; Kokubu, Yoko; Niwa, Masakazu; Watanabe, Takahiro; Saiga, Atsushi; Sueoka, Shigeru; Komatsu, Tetsuya; et al.

JAEA-Review 2018-020, 46 Pages, 2019/01

JAEA-Review-2018-020.pdf:1.25MB

This report is a plan of research and development (R&D) on geosphere stability for long-term isolation of high-level radioactive waste (HLW) in Japan Atomic Energy Agency, in fiscal year 2018. The objectives and contents in fiscal year 2018 are described in detail based on the outline of 7 years plan (fiscal years 2015-2021). Background of this research is clarified with the necessity and the significance for site investigation and safety assessment, and the past progress in this report. In addition, the plan framework is structured into the following categories: (1) Development and systematization of investigation techniques, (2) Development of models for long-term estimation and effective assessment, (3) Development of dating techniques.

JAEA Reports

Annual report for research on geosphere stability for long-term isolation of radioactive waste in fiscal year 2016

Ishimaru, Tsuneari; Yasue, Kenichi*; Asamori, Koichi; Kokubu, Yoko; Niwa, Masakazu; Watanabe, Takahiro; Yokoyama, Tatsunori; Fujita, Natsuko; Saiga, Atsushi; Shimizu, Mayuko; et al.

JAEA-Research 2018-008, 83 Pages, 2018/12

JAEA-Research-2018-008.pdf:11.43MB

This annual report documents the progress of research and development (R&D) in the 2nd fiscal year during the JAEA 3rd Mid- and Long-term Plan (fiscal years 2015-2021) to provide the scientific base for assessing geosphere stability for long-term isolation of the high-level radioactive waste. The planned framework is structured into the following categories: (1) Development and systematization of investigation techniques, (2) Development of models for long-term estimation and effective assessment, (3) Development of dating techniques. In this paper, the current status of R&D activities with previous scientific and technological progress is summarized.

Journal Articles

TEF beam window design and evaluation of structural integrity

Obayashi, Hironari; Takei, Hayanori; Wan, T.; Kogawa, Hiroyuki; Iwamoto, Hiroki; Sasa, Toshinobu

JPS Conference Proceedings (Internet), 8, p.041002_1 - 041002_7, 2015/09

Journal Articles

Mesoscopic structures of vermiculite and weathered biotite clays in suspension with and without cesium ions

Motokawa, Ryuhei; Endo, Hitoshi*; Yokoyama, Shingo*; Ogawa, Hiroki*; Kobayashi, Toru; Suzuki, Shinichi; Yaita, Tsuyoshi

Langmuir, 30(50), p.15127 - 15134, 2014/12

 Times Cited Count:11 Percentile:46.5(Chemistry, Multidisciplinary)

Journal Articles

Monopole-driven shell evolution below the doubly magic nucleus $$^{132}$$Sn explored with the long-lived isomer in $$^{126}$$Pd

Watanabe, H.*; Lorusso, G.*; Nishimura, Shunji*; Otsuka, T.*; Ogawa, K.*; Xu, Z. Y.*; Sumikama, Toshiyuki*; S$"o$derstr$"o$m, P.-A.*; Doornenbal, P.*; Li, Z.*; et al.

Physical Review Letters, 113(4), p.042502_1 - 042502_6, 2014/07

 Times Cited Count:13 Percentile:26.29(Physics, Multidisciplinary)

Journal Articles

Study on the structural integrity of beam window for TEF target

Takei, Hayanori; Obayashi, Hironari; Iwamoto, Hiroki; Kogawa, Hiroyuki; Sasa, Toshinobu

Proceedings of 11th International Topical Meeting on Nuclear Applications of Accelerators (AccApp '13), p.311 - 316, 2014/05

The objective of this study is to evaluate the feasibility of a designed beam window of TEF target by the numerical analysis with a 3D model. The analysis was performed by considering (1) the peak current density and shape of the incident beam, (2) the thermal-fluid behaviour of LBE around the beam window as a function of the flow rate and inlet temperature, (3) the material and the thickness of the beam window, (4) the structural strength of the beam window. In the reference case, the peak current density and the profile of the proton beam were 20 $$mu$$A/cm$$^{2}$$ and a Gaussian shape, respectively. The flow rate of LBE and temperature at the inlet were 1 $$ell/sec$$ and 350 $$^{circ}$$C. The material of a beam window was type SUS316 stainless steel with the 2 mm thick. In this reference case, the maximum velocity of LBE and the maximum temperature located at the top of the beam window were about 1.2 m/sec and 477 $$^{circ}$$C. By increasing the flow rate of LBE up to 4 $$ell/sec$$, the maximum temperature of a beam window was reduced around 420 $$^{circ}$$C. The maximum tresca stress was 190 MPa, which was observed at the center on the outside surface of a beam window. The analyzed stress in the reference case was lower than the tolerance level of the stress strength of the material, and hence the feasibility of a designed beam window was confirmed.

Journal Articles

Green hydrogen production by using nuclear energy for hydrogen steelmaking

Ogawa, Masuro; Kasahara, Seiji; Inagaki, Yoshiyuki; Noguchi, Hiroki

"Gurin Enerugi Seitetsu Kenkyukai" Seika Hokokusho, p.4 - 44, 2012/03

Green hydrogen production by using nuclear energy for hydrogen steelmaking is one of the candidates of CO$$_{2}$$ emission reduction. Very high temperature reactor (VHTR) is the most appropriate reactor type among generation IV nuclear reactors. In Japan, a large-scale national R&D project for nuclear steelmaking had been carried out during 1973-1980. The process was: reforming of bottom residue of oil distillation using nuclear heat to produce reducing gas: direct reduction of ore to produce iron by the reforming gas. R&D of 6 themes had been conducted: total systems, high temperature heat exchangers, very high temperature resistant alloy, high temperature thermal insulation materials, reforming gas generators, and shaft furnaces to produce direct reduced iron. A new type nuclear steelmaking is proposed now by the requirement of technology and a new problem of CO$$_{2}$$ reduction. The largest change is the shift of reducing gas to hydrogen. R&D of VHTR continues since the national project and thermochemical water splitting IS (iodine-sulfur) process has been studied for hydrogen production in Japan Atomic Energy Agency. Conceptual design of a hydrogen steelmaking process applying VHTR-IS process and direct reduction steelmaking with hydrogen had been carried out in Green Energy Steelmaking Research Group in Iron and Steel Institute of Japan during 2008-2012.

Journal Articles

Microscopic structures of tri-$$n$$-butyl phosphate/$$n$$-octane mixtures by X-ray and neutron scattering in a wide $$q$$ range

Motokawa, Ryuhei; Suzuki, Shinichi; Ogawa, Hiroki*; Antonio, M. R.*; Yaita, Tsuyoshi

Journal of Physical Chemistry B, 116(4), p.1319 - 1327, 2012/02

 Times Cited Count:23 Percentile:36.11(Chemistry, Physical)

Journal Articles

Application of thermochemical hydrogen production iodine-sulfur process to shaft furnace direct reduction ironmaking process

Kasahara, Seiji; Kubo, Shinji; Noguchi, Hiroki; Ohashi, Hirofumi; Ogawa, Masuro

Proceedings of International Symposium on Ironmaking for Sustainable Development 2010, p.84 - 87, 2010/01

Iodine-sulfur process (IS process) is a hydrogen production process from water by using heat around 900 $$^{circ}$$C without CO$$_{2}$$ production. An ironmaking plant combining IS process and hydrogen direct reduction shaft furnace is proposed. Heat/mass balance analysis of the plant was carried out by flowsheet calculation. Comparison of the process with a conventional blast furnace plant and a direct reduction process using reformed natural gas was carried out. Though total heat input to the process was greatest in these plants, CO$$_{2}$$ production of the hydrogen plant was 8 % of the blast furnace plant.

Journal Articles

A Preliminary comprehensive dynamic analysis of the typical FaCT scenarios with JSFR and related fuel cycle facilities

Shiotani, Hiroki; Ono, Kiyoshi; Ogawa, Takashi; Koma, Yoshikazu; Kawaguchi, Koichi

Proceedings of 2009 International Congress on Advances in Nuclear Power Plants (ICAPP '09) (CD-ROM), p.9419_1 - 9419_10, 2009/05

Dynamic analyses of the typical FR deployment scenarios with JSFR and related fuel cycle facilities developed in FaCT project was conducted. The total cash out-flows and the average electricity generation costs to 22nd century were calculated to seize the long-term economics as well as the material compositions in the nuclear facilities, the quantities of radioactive wastes generations. Several cash out-flow peaks and radioactive waste generation peaks were found because of the decommissioning and construction of the nuclear power plants and reprocessing plants for LWR spent fuel. Then, different breeding ratio, single/dual-purpose reprocessing plant, and with/without Am-Cm recycling were compared. For example, the exploration of the optimal breeding ratio between B.R. =1.1 and 1.2 for the start up stage FR was considered to be reasonable from the analysis.

Journal Articles

Temperature-dependent nano-scale dynamics of PVA physical gel

Takahashi, Nobuaki; Nishida, Koji*; Inoue, Rintaro*; Ogawa, Hiroki*; Kanaya, Toshiji*; Nagao, Michihiro*

NSL News Letter, 2007-4, p.155 - 157, 2007/04

We have studied dynamics of poly(vinyl alcohol) (PVA) gel in a mixture of deuterated dimethyl sulfoxide (DMSO-d$$_{6}$$) and D$$_{2}$$O (60/40 by volume) during heating process from 25$$^{circ}$$C to 80$$^{circ}$$C using neutron spin-echo (NSE) techniques.

JAEA Reports

Analyses of neutronic characteristics of STACY heterogeneous cores composed of 6wt%-enriched uranyl nitrate solution containing gadolinium and 1.5cm-lattice-pitch fuel pins

Izawa, Kazuhiko; Aoyama, Yasuo; Sono, Hiroki; Ogawa, Kazuhiko; Yanagisawa, Hiroshi

JAEA-Technology 2007-001, 40 Pages, 2007/02

JAEA-Technology-2007-001.pdf:2.73MB

A series of critical experiments is conducted in FY 2006 using a heterogeneous core of the Static Experiment Critical Facility (STACY) in the Japan Atomic Energy Agency (JAEA). In the experiment, the core is composed of uranyl nitrate solution ($$^{235}$$U enrichment 6wt%) containing soluble poison (gadolinium) and 333 pins of uranium dioxide ($$^{235}$$U enrichment 5wt%) loaded at a latticepitch of 1.5cm. Prior to the experiment, the following neutronic characteristics were analyzed to assess safety of the core and operation parametor limits: criticality, reactivity and reactor shutdown margins. In the analyses, a Monte Carlo code, MVP, and a neutronics code system, SRAC, were used with an evaluated nuclear data library, JENDL-3.3. From these analyses, it was confirmed that the reactor shutdown margins would comply with the safety criteria under all conditions of the fuel used in the experiments. Simplified formulas for criticality and reactivity were also evaluated based on the analyzed values which are utilized to confirm the operation parameter limits during operations of the core.

Journal Articles

Operation and management of STACY experiment using Pseudo-FPs-doped fuel

Izawa, Kazuhiko; Seki, Masakazu; Hirose, Hideyuki; Kaminaga, Jota; Aoyama, Yasuo; Yoshida, Hiroshi; Sono, Hiroki; Ogawa, Kazuhiko; Sakuraba, Koichi

UTNL-R-0453, p.9_1 - 9_10, 2006/03

no abstracts in English

Journal Articles

Neutron spin-echo studies on poly(vinyl alcohol) gels during melting process

Takahashi, Nobuaki; Nishida, Koji*; Tsubouchi, Tsuyoshi*; Ogawa, Hiroki*; Inoue, Rintaro*; Kanaya, Toshiji*; Nagao, Michihiro*

ISSP Activity Report on Neutron scattering Research; Experimental Reports (CD-ROM), 13, 2 Pages, 2006/00

no abstracts in English

Journal Articles

Report on investigation of unexpected reactor shutdown of the TRACY facility; Cause and countermeasures

Sono, Hiroki; Tsukamoto, Michio; Aizawa, Eiju; Takeuchi, Masaki; Fukaya, Yuji; Iseda, Hirokatsu*; Ogawa, Kazuhiko; Sakuraba, Koichi; Tonoike, Kotaro

UTNL-R-0446, p.3_1 - 3_10, 2005/03

This report describes an investigation and countermeasures of the unexpected reactor shutdown of the TRACY facility at the Japan Atomic Energy Research Institute on June 17, 2004, caused by a malfunction of a safety rod of the facility. As a result of the investigation, the principal cause of the malfunction was ascribed to a drop in attraction of a binding magnet for the safety rod by a small piece of a foreign material. The following countermeasures were taken against its recurrence: prevention of incidence of foreign materials, reduction of the chance of intrusion of foreign materials into devices, and certain detection of foreign materials. This lesson learned through the experience is expected to be shared with the staffs concerned in reactor operation and maintenance.

JAEA Reports

Feasibility Study on Commercialization of Fast Breeder Reactor Cycle Systems Interim Report of Phase II; Technical Study Report for Reactor Plant Systems

Konomura, Mamoru; Ogawa, Takashi; Okano, Yasushi; Yamaguchi, Hiroyuki; Murakami, Tsutomu; Takaki, Naoyuki; Nishiguchi, Youhei; Sugino, Kazuteru; Naganuma, Masayuki; Hishida, Masahiko; et al.

JNC-TN9400 2004-035, 2071 Pages, 2004/06

JNC-TN9400-2004-035.pdf:76.42MB

The attractive concepts for Sodium-, lead-bismuth-, helium- and water-cooled FBRs have been created through using typical plant features and employing advanced technologies. Efforts on evaluating technological prospects of feasibility have been paid for these concepts. Also, it was comfirmed if these concepts satisfy design requierments of capability and performance presumed in the feasibilty study on commertialization of Fast Breeder Reactor Systems. As results, it was concluded that the selection of sodium-cooled reactor was most rational for practical use of FBR technologies in 2015.

Journal Articles

High performance tokamak experiments with a ferritic steel wall on JFT-2M

Tsuzuki, Kazuhiro; Kimura, Haruyuki; Kawashima, Hisato; Sato, Masayasu; Kamiya, Kensaku; Shinohara, Koji; Ogawa, Hiroaki; Hoshino, Katsumichi; Bakhtiari, M.; Kasai, Satoshi; et al.

Nuclear Fusion, 43(10), p.1288 - 1293, 2003/10

 Times Cited Count:34 Percentile:24.33(Physics, Fluids & Plasmas)

no abstracts in English

JAEA Reports

Conceptual design study of the actinide recycle reactor; Study of core structural with ductless fuel assemblies

Ogawa, Shinta; Hayafune, Hiroki; Tozawa, Katsuhiro; Ichimiya, Masakazu; Hayashi, Hideyuki; Mukaibo, Ryuichi

PNC-TN9430 96-007, 354 Pages, 1996/07

PNC-TN9430-96-007.pdf:13.67MB

It is required enhanced core safety characteristics, recycle cost reduction, mitigation of risk to the environment and nuclear non proliferation for the fast reactor at the commercial use age of actinide recycle. The ductless fuel assembly, which has no wrapper tube, is promising for these requirements. In this study, the thermal hydraulic and mechanical characteristics of the ductless fuel core are evaluated for 600MWe MOX core with high burnup and long operating cycle length, and conceptual structure of the ductless fuel assembly core was established. The results of the study are summarized as follows; (1)Structural of Ductless Fuel Assembly. Conceptual design of components of the ductless fuel assembly, e.g. grid spacer, tie rod, upper shielding, lower nozzle and mechanical hold down spring, were performed and conceptual structure was established. Detail study of fuel pin bundle stiffness are required in the following design study. (2)Thermo-hydraulic Characteristics of Ductless Core and Ductless Fuel Assembly. The bypass flow rate strongly depends on the gap between core region and core barrel. For this bypass flow, it is found that thermal hydraulic feasibility is expected when the gaps between core region and core barrel are decreased($$<$$1mm). Since the core flow distribution is uniform, a coolant temperature distribution depend on the power distribution into core region. For fuel assembly, if the gaps between fuel assemblies are enlarged, the maximum sodium temperature increases (20$$^{circ}$$C/mm), therefore a proper gap design are needed. (3)Mechanical Characteristics of Ductless Core. The seismic safety of ductless core, in which a mechanical hold down are used, is assured. To decrease the impact force at spacer grid, however, some considerations on the grid design is necessary to avoid buckling. (4)Thermohydraulic Safety Characteristics. The maximum sodium temperatures are roughly evaluation under the condition of natural circulation and coolant ...

JAEA Reports

JFY 1995 Progress report of the development on the actinide recycle test reactor(ARTR)

Kasai, Shigeo; Tozawa, Katsuhiro; Akatsu, Minoru; Ogawa, Shinta; Watanabe, Ichiro; Hayafune, Hiroki; Naganuma, Masayuki; Ichimiya, Masakazu; Hayashi, Hideyuki; Mukaibou, Ryuichi

PNC-TN9430 96-004, 152 Pages, 1996/07

PNC-TN9430-96-004.pdf:6.15MB

Authors are studying the Actinide Recycle Fast Breeder Reactor (named ARFBR in this paper), which contribute to the reduction of burdens to environments and to enhance the capability to prevent the nuclear proliferation as the entire nuclear recycle system (named Advanced Fuel Recycling FBR system (AFRFS) in this paper), and also investigating the ARTR for developing the ARFBR. The investigation of the ARTR consists of the design study of the ARTR and R&Ds of key technology existing in ARTR concept. The conceptual design study of the ARTR is planed to be conducted for 2 years from 1995 to 1996 as first stage. 1995's design study have been performed with drawing over all plant concept with supposing various tests in reactor and usage of reactor. Followings are distinctive feature of 1995's design study. (1)Maximum reactor power is 400MWt with about 1.6m diameter irradiation (burning) cores, which are designed to be operated up to 150GWd/t as average burn up. Maximum core diameter is about 2.5m for low power nuclear physics tests which are designed to be able to estimate characteristics of large scale core by using the test results. (2)Mixed oxide (MOX) and Mixed nitride (MN) core is designed respectively to be able to be used for static nuclear physics test, for nuclear and thermal transient test, and for full power irradiation or burning test. Each core is designed to terminate ATWS events passively, with using GEM for MOX core and with using spectral adjustment for MN core. (3)Fuel assembly is employed ductless type which is a promising candidate for the ARFBR. Sizing of a fuel assembly is determined in basis on MOX fuel design because MOX fuel pin length covers MN fuel pin which accommodates lesser FP gases because of its lower temperature. Fuel assembly is managed to be held by hydraulic force in case of freeing mechanical stopper by requirement of testability. (4)Reactor assembly is designed based on so called Head Access Loop Type Reactor. Main changes ...

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