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Journal Articles

Public acceptance of nuclear waste disposal sites; A Decision-making process utilising the "veil of ignorance" concept

Yokoyama, Miki*; Onuma, Susumu*; Osawa, Hideaki; Otomo, Shoji*; Hirose, Yukio*

Humanities & Social Sciences Communications (Internet), 10(1), p.623_1 - 623_10, 2023/09

This study demonstrates that a decision-making process utilising "the veil of ignorance" concept, defined in process terms as beginning from a blank slate encompassing the entire country as potential sites and shortlisting candidate sites based on scientific (geological) safety, promotes public acceptance of siting a repository for the geological disposal of high-level radioactive waste and fosters procedural fairness.

Journal Articles

Can the decision way by the "veil of ignorance" promote public acceptance?; A Hypothetical scenario survey about selecting a candidate site for geological disposal of high level radioactive waste in Japan

Onuma, Susumu*; Hirose, Yukio*; Osawa, Hideaki; Otomo, Shoji*; Yokoyama, Miki*

Nihon Risuku Kenkyu Gakkai Dai-31-Kai Nenji Taikai Koen Rombunshu (USB Flash Drive), 6 Pages, 2018/11

One reason of not being decided a candidate site for geological disposal of high level radioactive waste is that strong opposition occurs once a name of the place where residents live even if they can understand the necessity of the facility. Therefore, we hypothesized that people would accept the decision if they consent the decision way beforehand under the condition everyone can be involved in as a potential interest party (the ignorance of veil). A hypothetical scenario survey was conducted in Japan. The results revealed that the decision way and the decision under the ignorance of veil were likely to be accepted more than the current decision way even if they became candidate site. Moreover, the decision way under the ignorance of veil was evaluated higher procedural fairness, while the current way was low procedural fairness.

Journal Articles

Research and development roadmap for reactor physics 2017; Future of reactor physics projected by next generations

Yamamoto, Akio*; Chiba, Go*; Kirimura, Kazuki*; Miki, Yosuke*; Yokoyama, Kenji

Nihon Genshiryoku Gakkai-Shi ATOMO$$Sigma$$, 60(4), p.241 - 245, 2018/04

no abstracts in English

Journal Articles

R&D status on water cooled ceramic breeder blanket technology

Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.

Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10

 Times Cited Count:20 Percentile:84.35(Nuclear Science & Technology)

The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.

Journal Articles

Progress of manufacturing and quality testing of the ITER divertor outer vertical target in Japan

Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Mori, Kensuke; Enoeda, Mikio

Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03

The outer vertical targets of the ITER divertor are procured by the Japan Domestic Agency, JADA. Manufacturing a full-scale prototype of the half cassette which consists of 11 plasma facing units and a steel support structure has been started in Japan. JADA has greatly improved the success rate of the joint between the plasma-facing materials and heat-sink materials, in consequence of R&D on joint technology and quality control. JADA solved problems of quality control of the joint interface by an improved system of infrared thermography inspection, which provides quick feedback during the manufacturing process about the presence of defect in the joint. This paper reports on the achievements and the clarifications of technical and quality issues for the manufacture of the divertor components to be supplied by Japan.

Journal Articles

Characterization of edge radial electric field structures in the large helical device and their viability for determining the location of the plasma boundary

Kamiya, Kensaku; Ida, Katsumi*; Yoshinuma, Mikiro*; Suzuki, Chihiro*; Suzuki, Yasuhiro*; Yokoyama, Masayuki*; LHD Experimental Group*

Nuclear Fusion, 53(1), p.013003_1 - 013003_9, 2013/01

 Times Cited Count:17 Percentile:60.32(Physics, Fluids & Plasmas)

This paper provides and proposes a new technique to determine the location of the LCFS that is based on a characterization of the $$E$$$$_{r}$$ structure derived from CXS measurements in the LHD. We found that the spatial derivative in the $$E$$$$_{r}$$ structure had the local maximum value at the region very near, or possibly outside the vacuum LCFS location of vacuum magnetic field at the outer midplane in the low $$beta$$ plasma.

Journal Articles

A Study on flow field of purge gas for tritium transfer through breeder pebble bed in fusion blanket

Seki, Yohji; Ezato, Koichiro; Yokoyama, Kenji; Enoeda, Mikio; Kubota, Jinichi*; Sakamoto, Kensaku

Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12

Japan Atomic Energy Agency has been performing R&D and design of a blanket module of a nuclear fusion reactor. Pebbles of a ceramic tritium breeder are packed in a container of the blanket. Helium purge gas is applied as a transport fluid in a tritium recovery system. Prediction of the flow phenomena with a tritium transfer is important for designs of the container. A purpose of our research is to establish and verify a method for a prediction of the flow in the pebble bed. In this study, pressure drops of the helium purge gas through the pebble bed were measured up to 100 L/min of flow rate. Reliability of prediction ability of the pressure drop was validated by this experiment within the flow rate which is less than 40 L/min. A numerical simulation for the flow field through the pebble bed also has been performed. Consequently, the velocity distributions are quantitatively and qualitatively obtained at near the wall and the center region in the pebble bed.

Journal Articles

Development of the plasma facing components in Japan for ITER

Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Enoeda, Mikio

Fusion Engineering and Design, 87(5-6), p.845 - 852, 2012/08

 Times Cited Count:17 Percentile:78.55(Nuclear Science & Technology)

After the successful completion of the prequalification activity for ITER divertor procurement, Japanese Domestic Agency (JADA) and ITER Organization (IO) have entered into the procurement arrangement of divertor outer vertical target (OVT) in June 2009. In accordance with the arrangement, JADA has started to manufacture an OVT full-scale prototype in order to pick out and solve technical and quality issues, then to establish a rational manufacturing process toward the start of the series of production of the OVT components to be installed in tokamak. This paper presents the overview of JADA's activity on the divertor outer target procurement and also procurement schedule will be presented.

Journal Articles

Recent status of fabrication technology development of water cooled ceramic breeder test blanket module in Japan

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

Fusion Engineering and Design, 86(9-11), p.2265 - 2268, 2011/10

 Times Cited Count:5 Percentile:38.79(Nuclear Science & Technology)

As one of the most important fabrication technologies of the WCCB TBM, Hot Isostatic Pressing (HIP) joining technology was selected to fabricate the first wall with built-in cooling channel structure made of reduced activation martensitic/ferritic steel, F82H. By using developed HIP technology, a real scale TBM first wall mockup was successfully fabricated. High heat flux test of the fabricated mockup showed the feasibility to with the equivalent conditions of the WCCB TBM operation. The breeder pebble box was successfully fabricated with thin wall cooling pipes and thin plate sleds by Laser welding. With respect to the side walls with built in cooling channels were also fabricated using drilling technology. Assembling of the first wall and side walls is one of the critical fabrication processes of the fabrication of the TBM structure. By using a F82H first wall mockup and side wall mockups, assembling process was demonstrated successfully by Electron Beam welding.

Journal Articles

Deuterium concentration of co-deposited carbon layer produced at gap of wall tiles

Nobuta, Yuji*; Yokoyama, Kenji; Kanazawa, Jun*; Yamauchi, Yuji*; Hino, Tomoaki*; Suzuki, Satoshi; Ezato, Koichiro; Enoeda, Mikio; Akiba, Masato

Journal of Nuclear Materials, 417(1-3), p.607 - 611, 2011/10

 Times Cited Count:2 Percentile:18.37(Materials Science, Multidisciplinary)

Journal Articles

Numerical simulation of turbulent flow of coolant in a test blanket module of nuclear fusion reactor

Seki, Yohji; Onishi, Yoichi*; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ozu, Akira; Ezato, Koichiro; Tsuru, Daigo; Suzuki, Satoshi; Yokoyama, Kenji; et al.

Progress in Nuclear Science and Technology (Internet), 2, p.139 - 142, 2011/10

R&D of a test blanket module (TBM) with a water-cooled solid breeder has been performed for ITER. For our design, the temperature of a coolant pressurized up to 15 MPa is designed as 598 K in an outlet of the TBM, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. The Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the pressure drop and flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mock-up.

Journal Articles

Integrated transport simulation of LHD plasmas using TASK3D

Wakasa, Arimitsu*; Fukuyama, Atsushi*; Murakami, Sadayoshi*; Miki, Masayuki*; Yokoyama, Masayuki*; Sato, Masahiko*; Toda, Shinichiro*; Funaba, Hisamichi*; Tanaka, Kenji*; Ida, Katsumi*; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2011/03

Journal Articles

Non-destructive examination with infrared thermography system for ITER divertor components

Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Enoeda, Mikio; Mori, Seiji

Fusion Engineering and Design, 85(7-9), p.1451 - 1454, 2010/12

 Times Cited Count:17 Percentile:74.17(Nuclear Science & Technology)

Japan Atomic Energy Agency (JAEA) is willing to procure the outer Vertical Targets (VT) in cooperation with ITER Organization. In advance of the start of the procurement, the JAEA has to first demonstrate its technical capability to carry out the procurement. This is achieved via the successful manufactures and quality tests of VT Qualification Prototypes. Non-Destructive Examination (NDE) with the infrared thermography is required as one of the quality tests to detect the defect in the CFC monoblock, and between the CFC/OFCu. In this research and development, the Facility of Infrared NDE for Divertor (FIND) has been built by the JAEA. The FIND successfully detects the position and the magnitude of the integrated defect in the CFC and in the bonding of CFC/OFCu. The infrared NDE system established in the JAEA contributes to keeping the quality of the ITER-divertor.

Journal Articles

Thermo-hydraulic testing and integrity of ITER test blanket module (TBM) first wall mock-up in JAEA

Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio

Fusion Engineering and Design, 85(7-9), p.1255 - 1260, 2010/12

 Times Cited Count:12 Percentile:62.45(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Numerical simulation of turbulent flow of coolant in a test blanket module of nuclear fusion reactor

Seki, Yohji; Onishi, Yoichi*; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ozu, Akira; Ezato, Koichiro; Tsuru, Daigo; Suzuki, Satoshi; Yokoyama, Kenji; et al.

Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 4 Pages, 2010/10

R&D of a test blanket module (TBM) with a water-cooled solid breeder has been performed for ITER. For our design, the temperature of a coolant pressurized up to 15 MPa is designed as 598 K in an outlet of the TBM, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. The Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the pressure drop and flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mockup.

Journal Articles

Overview of the R&D activities of water cooled ceramic breeder blanket

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

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), p.645 - 649, 2010/05

This paper overviews the research and development activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing development of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m$$^{2}$$, 80 cycles with the coolant condition as DEMO, 15 MPa 300 $$^{circ}$$C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been almost completed.

JAEA Reports

Characteristics of water flow distribution in TBM side wall

Yoshikawa, Akira; Tanigawa, Hisashi; Seki, Yohji; Hirose, Takanori; Tsuru, Daigo; Ezato, Koichiro; Yokoyama, Kenji; Nishi, Hiroshi; Suzuki, Satoshi; Tanzawa, Sadamitsu; et al.

JAEA-Technology 2009-077, 23 Pages, 2010/03

JAEA-Technology-2009-077.pdf:2.62MB

In the side wall of TBM, parallel flow channels are considered. In the cooling channels structure, the flow distribution probably arises from the pressure drop in the channels. The purpose of this study is to clarify the water flow distribution in the side wall and design the cooling channels structure so that structural material of the side wall can be kept under the allowable temperature. The structural material for assumed flow rates and the flow distribution were estimated, and then the cooling channels structure was designed. The design was verified using the mockup made of the vinyl chloride pipe. For the verified design, the mockup made of F82H is manufactured, and the water flow distribution and the pressure drop were measured. It was found that the heat removal capability was sufficient in this design. From these results, the design for the cooling channels structure in the side wall is established so that enough water flow to cool the structural material is kept.

Journal Articles

Recent activities related to the development of the plasma facing components for the ITER and fusion DEMO plant

Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji; Hirose, Takanori; Mori, Seiji; Enoeda, Mikio

Physica Scripta, T138, p.014003_1 - 014003_5, 2009/12

 Times Cited Count:3 Percentile:26.35(Physics, Multidisciplinary)

JAEA is going to procure the whole ITER divertor outer vertical target. The qualification of the ITER divertor components has been started to validate the technical capability of the domestic agencies. JAEA has developed vertical target qualification prototypes which cover most of the critical technical issues toward the series of production. The prototypes have been high heat flux tested in Efremov institute, and showed sufficient durability at 20 MW/m$$^{2}$$. JAEA has successfully obtained the certification from ITER organization. Development of a breeding blanket is one of the most important issues to realize DEMO plant. The engineering testing of the Test Blanket Module (TBM) in ITER is a key milestone toward the DEMO blanket. R&D's on the water-cooled solid breeder blanket has been performed in JAEA. A full-scale length TBM first wall mock-up made of F82H, has been developed by using HIP bonding technique. This mock-up showed sound thermal performance in the high heat flux test.

Journal Articles

Anionic fluoro complex of element 105, Db

Kasamatsu, Yoshitaka*; Toyoshima, Atsushi; Asai, Masato; Tsukada, Kazuaki; Li, Z.; Ishii, Yasuo; Tome, Hayato*; Sato, Tetsuya; Kikuchi, Takahiro; Nishinaka, Ichiro; et al.

Chemistry Letters, 38(11), p.1084 - 1085, 2009/10

 Times Cited Count:14 Percentile:49.01(Chemistry, Multidisciplinary)

We report on the characteristic anion-exchange behavior of the superheavy element dubnium (Db) with atomic number Z = 105 in HF/HNO$$_{3}$$ solution at the fluoride ion concentration [F$$^{-}$$] = 0.003 M. The result clearly demonstrates that the fluoro complex formation of Db is significantly different from that of the group-5 homologue Ta in the 6th period of the periodic table while the behavior of Db is similar to that of the lighter homologue Nb in the 5th period.

Journal Articles

Dynamics of ion internal transport barrier in LHD heliotron and JT-60U tokamak plasmas

Ida, Katsumi*; Sakamoto, Yoshiteru; Yoshinuma, Mikiro*; Takenaga, Hidenobu; Nagaoka, Kenichi*; Hayashi, Nobuhiko; Oyama, Naoyuki; Osakabe, Masaki*; Yokoyama, Masayuki*; Funaba, Hisamichi*; et al.

Nuclear Fusion, 49(9), p.095024_1 - 095024_9, 2009/09

 Times Cited Count:29 Percentile:72.19(Physics, Fluids & Plasmas)

Dynamics of ion internal transport barrier (ITB) formation and impurity transport both in the Large Helical Device (LHD) heliotron and JT-60U tokamak are described. Significant differences between heliotron and tokamak plasmas are observed. The location of the ITB moves outward during the ITB formation regardless of the sign of magnetic shear in JT-60U and the ITB becomes more localized in the plasma with negative magnetic shear. In LHD, the low Te/Ti ratio ($$<$$ 1) of the target plasma for the high power heating is found to be necessary condition to achieve the ITB plasma and the ITB location tends to expand outward or inward depending on the condition of the target plasmas. Associated with the formation of ITB, the carbon density tends to be peaked due to inward convection in JT-60U, while the carbon density becomes hollow due to outward convection in LHD. The outward convection observed in LHD contradicts the prediction by neoclassical theory.

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