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Journal Articles

Sorption of Eu$$^{3+}$$ on Na-montmorillonite studied by time-resolved laser fluorescence spectroscopy and surface complexation modeling

Sasaki, Takayuki*; Ueda, Kenyo*; Saito, Takumi; Aoyagi, Noboru; Kobayashi, Taishi*; Takagi, Ikuji*; Kimura, Takaumi; Tachi, Yukio

Journal of Nuclear Science and Technology, 53(4), p.592 - 601, 2016/04

 Times Cited Count:9 Percentile:75.69(Nuclear Science & Technology)

The influences of pH and the concentrations of Eu$$^{3+}$$ and NaNO$$_{3}$$ on the sorption of Eu$$^{3+}$$ to Na-montmorillonite were investigated through batch sorption measurements and time-resolved laser fluorescence spectroscopy (TRLFS). The pH had a little effect on the distribution coefficients (Kd) in 0.01 M NaNO$$_{3}$$, whereas the Kd strongly depended on pH at 1 M NaNO$$_{3}$$. A cation exchange model combined with a one-site non-electrostatic surface complexation model was successfully applied to the measured Kd. The TRLFS spectra of Eu$$^{3+}$$ sorbed were processed by parallel factor analysis (PARAFAC), which corresponded to one outer-sphere (factor A) and two inner-sphere (factor B and C) complexes. It turned out that factors A and B correspond to Eu$$^{3+}$$ sorbed by ion exchange sites and inner-sphere complexation with hydroxyl groups of the edge faces, respectively. Factor C became dominant at relatively high pH and ionic strength and likely correspond to the precipitation of Eu(OH)$$_{3}$$ on the surface.

Journal Articles

2016 Professional Engineer (PE) test preparation course "Nuclear and Radiation Technical Disciplines"

Takahashi, Naoki; Yoshinaka, Kazuyuki; Harada, Akio; Yamanaka, Atsushi; Ueno, Takashi; Kurihara, Ryoichi; Suzuki, Soju; Takamatsu, Misao; Maeda, Shigetaka; Iseki, Atsushi; et al.

Nihon Genshiryoku Gakkai Homu Peji (Internet), 64 Pages, 2016/00

no abstracts in English

Journal Articles

Dismantlement of large fusion experimental device JT-60U

Ikeda, Yoshitaka; Okano, Fuminori; Sakasai, Akira; Hanada, Masaya; Akino, Noboru; Ichige, Hisashi; Kaminaga, Atsushi; Kiyono, Kimihiro; Kubo, Hirotaka; Kobayashi, Kazuhiro; et al.

Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(4), p.167 - 178, 2014/12

The JT-60U torus was disassembled so as to newly install the superconducting tokamak JT-60SA torus. The JT-60U used the deuterium for 18 years, so the disassembly project of the JT-60U was the first disassembly experience of a fusion device with radioactivation in Japan. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. The lessons learned from the disassembly project indicated that the cutting technologies and storage management of disassembly components were the key factors to conduct the disassembly project in an efficient way. After completing the disassembly project, efforts have been made to analyze the data for characterizing disassembly activities, so as to contribute the estimation of manpower needs and the radioactivation of the disassembly components on other fusion devices.

Journal Articles

Safe disassembly and storage of radioactive components of JT-60U torus

Ikeda, Yoshitaka; Okano, Fuminori; Hanada, Masaya; Sakasai, Akira; Kubo, Hirotaka; Akino, Noboru; Chiba, Shinichi; Ichige, Hisashi; Kaminaga, Atsushi; Kiyono, Kimihiro; et al.

Fusion Engineering and Design, 89(9-10), p.2018 - 2023, 2014/10

 Times Cited Count:2 Percentile:19.81(Nuclear Science & Technology)

Disassembly of the JT-60U torus was started in 2009 after 18-years D$$_{2}$$ operations, and was completed in October 2012. The JT-60U torus was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to D-D reactions. Since this work is the first experience of disassembling a large radioactive fusion device in Japan, careful disassembly activities have been made. About 13,000 components cut into pieces with measuring the dose rates were removed from the torus hall and stored safely in storage facilities by using a total wokers of 41,000 person-days during 3 years. The total weight of the disassembly components reached up to 5,400 tons. Most of the disassembly components will be treated as non-radioactive ones after the clearance verification under the Japanese regulation in future. The assembly of JT-60SA has started in January 2013 after this disassembly of JT-60U torus.

JAEA Reports

Research on engineering technology in the full-scale demonstration of EBS and operation technology for HLW disposal; Research report in 2012 (Joint research)

Nakatsuka, Noboru; Sato, Haruo; Tanai, Kenji; Nakayama, Masashi; Sawada, Sumiyuki*; Asano, Hidekazu*; Saito, Masahiko*; Yoshino, Osamu*; Tsukahara, Shigeki*; Hishioka, Sosuke*; et al.

JAEA-Research 2013-034, 70 Pages, 2014/01

JAEA-Research-2013-034.pdf:9.11MB

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 2012 (2011/2012) 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 2012. In fiscal year 2012, part of the equipments for emplacement of buffer material was produced and visualization test for water penetration in buffer material were carried out.

Journal Articles

Achievement of 500 keV negative ion beam acceleration on JT-60U negative-ion-based neutral beam injector

Kojima, Atsushi; Hanada, Masaya; Tanaka, Yutaka*; Kawai, Mikito*; Akino, Noboru; Kazawa, Minoru; Komata, Masao; Mogaki, Kazuhiko; Usui, Katsutomi; Sasaki, Shunichi; et al.

Nuclear Fusion, 51(8), p.083049_1 - 083049_8, 2011/08

 Times Cited Count:42 Percentile:88.1(Physics, Fluids & Plasmas)

Hydrogen negative ion beams of 490 keV, 3 A and 510 keV, 1 A have been successfully produced in the JT-60 negative ion source with three acceleration stages. These successful productions of the high-energy beams at high current have been achieved by overcoming the most critical issue, i.e., a poor voltage holding of the large negative ion sources with the grids of 2 m$$^{2}$$ for JT-60SA and ITER. To improve voltage holding capability, the breakdown voltages for the large grids was examined for the first time. It was found that a vacuum insulation distance for the large grids was 6-7 times longer than that for the small-area grid (0.02 m$$^{2}$$). From this result, the gap lengths between the grids were tuned in the JT-60 negative ion source. The modification of the ion source also realized a significant stabilization of voltage holding and a short conditioning time. These results suggest a practical use of the large negative ion sources in JT-60SA and ITER.

Journal Articles

Demonstration of 500 keV beam acceleration on JT-60 negative-ion-based neutral beam injector

Kojima, Atsushi; Hanada, Masaya; Tanaka, Yutaka*; Kawai, Mikito*; Akino, Noboru; Kazawa, Minoru; Komata, Masao; Mogaki, Kazuhiko; Usui, Katsutomi; Sasaki, Shunichi; et al.

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

Hydrogen negative ion beams of 490keV, 3A and 510 keV, 1A have been successfully produced in the JT-60 negative ion source with three acceleration stages. These successful productions of the high-energy beams at high current have been achieved by overcoming the most critical issue, i.e., a poor voltage holding of the large negative ion sources with the grids of $$sim$$ 2 m$$^{2}$$ for JT-60SA and ITER. To improve voltage holding capability, the breakdown voltages for the large grids was examined for the first time. It was found that a vacuum insulation distance for the large grids was 6-7 times longer than that for the small-area grid (0.02 m$$^{2}$$). From this result, the gap lengths between the grids were tuned in the JT-60 negative ion source. The modification of the ion source also realized a significant stabilization of voltage holding and a short conditioning time. These results suggest a practical use of the large negative ion sources in JT-60 SA and ITER.

JAEA Reports

Fuel and core design studies on metal fuel sodium-cooled fast reactor, 3; Joint research report for JFY2007&2008

Okano, Yasushi; Kobayashi, Noboru*; Ogawa, Takashi; Oki, Shigeo; Naganuma, Masayuki; Okubo, Tsutomu; Mizuno, Tomoyasu; Ogata, Takanari*; Ueda, Nobuyuki*; Nishimura, Satoshi*

JAEA-Research 2009-025, 105 Pages, 2009/10

JAEA-Research-2009-025.pdf:10.45MB

A metal fuel core has specific features on high heavy metal density, hard neutron spectrum, and efficient neutron utilization. Enlarged applicable design envelops would improve core performances and features: higher breeding ratio, compacted reactor core, and, smaller amount of Pu-fissile inventory. A joint study on "Reactor Core and Fuel Design of Metal Fuel Core of Sodium Cooled Fast Reactor" by Japan Atomic Energy Agency and Central Research Institute of Electric Power Industry has been conducted during Japanese fiscal years of 2007 and 2008. This report shows the results on (1) the study on applicable design ranges of metal fuel specifications, (2) the study on conceptual core designs for high breeding ratio, and (3) the safety study on metal fuel core designed in the Fast Reactor Cycle Technology Development (FaCT) Project.

Journal Articles

Energy spectra of bremsstrahlung X-rays emitted from an FRP insulator

Tanaka, Yutaka; Ikeda, Yoshitaka; Hanada, Masaya; Kobayashi, Kaoru; Kamada, Masaki; Kisaki, Masashi; Akino, Noboru; Yamano, Yasushi*; Kobayashi, Shinichi*; Grisham, L. R.*

IEEE Transactions on Plasma Science, 37(8), p.1495 - 1498, 2009/08

 Times Cited Count:1 Percentile:4.64(Physics, Fluids & Plasmas)

Voltage holding capability of the JT-60 negative ion source is limited by surface flashover on the FRP insulator. To improve the voltage holding capability of the ion source, the understanding of the surface flashover is required. In this study, electron energy is estimated by measuring the bremsstrahlung X-ray emitted from an FRP insulator. Energy spectra of X-ray were measured for 3 different positions and compared with those of the vacuum gap between electrodes. Near the anode, X-ray spectrum was dominated by the monoenergetic electron. Near the cathode, spectrum peak shifted to low energy compared with that near the anode. This result showed that a large amount of low energy electrons was generated on the surface of the FRP insulator near the cathode.

Journal Articles

Characteristics of voltage holding capability in multi-stage large electrostatic accelerator for fusion application

Kobayashi, Kaoru; Hanada, Masaya; Akino, Noboru; Sasaki, Shunichi; Ikeda, Yoshitaka; Takahashi, Masahiro*; Yamano, Yasushi*; Kobayashi, Shinichi*; Grisham, L. R.*

IEEE Transactions on Dielectrics and Electrical Insulation, 16(3), p.871 - 875, 2009/06

 Times Cited Count:1 Percentile:12.71(Engineering, Electrical & Electronic)

Voltage holding capability of a 500kV, 22A three-stage electrostatic accelerator, where large-area grids of 0.28 m$$^{2}$$ and large FRP insulators of 1.8 m in diameter are used, was examined. High voltage was independently applied to each acceleration stage, where the voltage holding capabilities of 130 kV were obtained. To identify whether the breakdowns occur in the gaps between the grids or the FRP insulators, high voltages were applied to the accelerator with and without the grids. Breakdown voltages without grids, i.e., the FRP insulator itself reached 170 kV of design value for each stage. These results show that the breakdown voltage of the accelerator was mainly determined by the gaps between the large-area grids. In this paper, the influence of non-uniform electric field and multi-stage grids on the voltage holding capabilities was also discussed.

Journal Articles

A Design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

Kobayashi, Noboru; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari*

Journal of Power and Energy Systems (Internet), 3(1), p.126 - 135, 2009/00

Utilizing advantages of the metal fuel core to the mixed oxide fuel one, such as its higher breeding ratio and compact core size, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8${$}$, a core height of less than 150 cm, a maximum cladding temperature of 650$$^{circ}$$C, and a fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40.

Journal Articles

Advanced LWR concept with hard neutron spectrum (FLWR) for realizing flexible plutonium management

Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro; Kobayashi, Noboru

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

An advanced LWR concept with hard neutron spectrum (FLWR) has been proposed in order to ensure sustainable energy supply in the future based on the well-experienced LWR technologies. The FLWR is essentially a BWR-type reactor, in which the moderation of neutron in the core is reduced by use of the hexagonal-shaped fuel assemblies with the triangular-tight-lattice fuel rod configuration. The core design concept of FLWR is to realize effective and flexible utilization of uranium and plutonium resources by two stages, corresponding to the advancement of the fuel cycle technologies and related infrastructures. The core in the first stage of FLWR aims at intensive utilization and preservation of plutonium based on the experiences of the current LWR and MOX utilization, and the one in the second stage realizes sustainable multiple plutonium recycling with a high conversion ratio over 1.0. The present paper summarizes the recent core design studies of FLWR.

Journal Articles

TRU recycling in BWR type reactor of FLWR with hard spectrum

Okubo, Tsutomu; Nakano, Yoshihiro; Fukaya, Yuji; Kobayashi, Noboru; Uchikawa, Sadao

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 3 Pages, 2008/09

In order to ensure sustainable energy supply in the future based on the well-established LWR technologies, conceptual design studies on the innovative water reactor for flexible fuel cycle (FLWR) have been performed at JAEA. FLWR is a BWR type advanced LWR concept with the triangular tight-lattice core of uranium (U) and plutonium (Pu) mixed oxide (MOX) fuel rods. Accordingly, FLWR can achieve a high conversion ratio from U to Pu in the hard neutron spectrum core. This core characteristic is also suitable for recycling of Pu and/or the minor actinides (MA) based on the fuel recycling strategy. FLWR core consists of two concepts of HC-FLWR and RMWR with different conversion ratios. It has been confirmed that even in HC-FLWR with a lower conversion ratio around 0.85 TRU recycling with about 2% MA would be possible.

Journal Articles

FBR core concepts in the "FaCT" Project in Japan

Oki, Shigeo; Ogawa, Takashi; Kobayashi, Noboru; Naganuma, Masayuki; Kawashima, Katsuyuki; Maruyama, Shuhei; Mizuno, Tomoyasu; Tanaka, Toshihiko*

Proceedings of International Conference on the Physics of Reactors, Nuclear Power; A Sustainable Resource (PHYSOR 2008) (CD-ROM), 10 Pages, 2008/09

Conceptual design studies of sodium-cooled fast reactor core are performed in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan. The representative MOX fuel core and the metal fuel core exert excellent performances on safety and reliability, sustainability, economic competitiveness, and nuclear non-proliferation. This paper reviews their feature in terms of reactor physics, and describes recent progress in design studies. In the recent design studies, much interest has been taken in the fuel composition change in the transition stage from light water reactors to fast breeder reactors. The core flexibility is also shown to fulfil the refined objectives such as high breeding and an enhancement of non-proliferation property.

Journal Articles

Recent R&D activities of negative-ion-based ion source for JT-60SA

Ikeda, Yoshitaka; Hanada, Masaya; Kamada, Masaki; Kobayashi, Kaoru; Umeda, Naotaka; Akino, Noboru; Ebisawa, Noboru; Inoue, Takashi; Honda, Atsushi; Kawai, Mikito; et al.

IEEE Transactions on Plasma Science, 36(4), p.1519 - 1529, 2008/08

 Times Cited Count:8 Percentile:34.65(Physics, Fluids & Plasmas)

The JT-60SA N-NBI system is required to inject 10 MW for 100 s at 500 keV. Three key issues should be solved for the JT-60SA N-NBI ion source. One is to improve the voltage holding capability. Recent R&D tests suggested that the accelerator with a large area of grids may need a high margin in the design of electric field and a long time for conditioning. The second issue is to reduce the grid power loading. It was found that some beamlets were strongly deflected due to beamlet-beamlet interaction and strike on the grounded grid. The grids are to be designed by taking account of beamlet-beamlet interaction in three-dimensional simulation. Third is to maintain the D- production for 100 s. A simple cooling structure is proposed for the active cooled plasma grid, where a key is the temperature gradient on the plasma grid for uniform D- production. The modified N-NBI ion source will start on JT-60SA in 2015.

JAEA Reports

Breakdown location without beam acceleration in the JT-60U negative ion source

Kobayashi, Kaoru; Hanada, Masaya; Kamada, Masaki; Akino, Noboru; Sasaki, Shunichi; Ikeda, Yoshitaka

JAEA-Technology 2008-042, 25 Pages, 2008/06

JAEA-Technology-2008-042.pdf:4.16MB

Breakdown locations of a JT-60U negative ion source were investigated to improve the voltage holding capability. The accelerator is characterized by three acceleration stages with large grids 0.45 m $$times$$ 1.1 m and large FRP insulators 1.8 m in inner diameter. High voltages were applied to each acceleration stage independently. Voltage holding capabilities of each stage were almost the same, $$sim$$ 120-130 kV, which was lower than the design acceleration voltage of 167 kV. Then, in order to identify whether the breakdowns occur in the gaps between grids or on the surface of the FRP insulators, high voltages were also applied to the accelerator with the grids and their support flanges removed. The voltage holding capabilities of three FRP insulators rapidly achieved 167 kV. These results indicate that the breakdowns mainly occur in the gaps between the acceleration grids and/or their support flanges.

JAEA Reports

Thermal-hydraulic design of high conversion type core of FLWR

Kobayashi, Noboru; Onuki, Akira; Uchikawa, Sadao; Okubo, Tsutomu

JAEA-Research 2008-054, 145 Pages, 2008/05

JAEA-Research-2008-054.pdf:2.39MB

A thermal-hydraulic design of the high-conversion (HC) type core of the innovative water reactor for flexible fuel cycle (FLWR) was constructed under the natural circulation core cooling, in order to achieve that HC-FLWR core can be converted to a breeder type one. The criteria on the thermal-hydraulic design of HC-FLWR were that the average void fractions in the core was smaller than 50%, and that the critical power ratio was larger than 1.3. The criterion on void fractions was determined from the nuclear design of HC-FLWR. The length of the chimney and the settings of the inlet orifice are common in both types of cores. The coefficient of the lower tie-plate of the HC-FLWR core and the temperature of the feed water were parametrically changed. Consequently, the design criteria were satisfied by adopting the setting of the form loss coefficients of the lower tie-plate comparable to those of the current BWRs and by lowering the feed water temperature to 505 K.

Journal Articles

Current status of design technology on core thermal-hydraulic performance in FLWR

Onuki, Akira; Kobayashi, Noboru

Dai-45-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu, 1, p.3 - 4, 2008/05

no abstracts in English

Journal Articles

Study on enhanced performance sodium-cooled metal fuel core concepts by adopting advanced fuel and flexible design criteria

Kobayashi, Noboru; Ogawa, Takashi; Oki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari*

Proceedings of 16th International Conference on Nuclear Engineering (ICONE-16) (CD-ROM), 9 Pages, 2008/05

The metal fuel core is superior to the mixed oxide fuel core because of its higher breeding ratio and compact core size resulting from neutron economics, hard neutron spectrum, and high content of heavy metal nuclides. Utilizing the advantage of the metal fuel core, conceptual sodium-cooled fast breeder reactor designs have been pursued for the attractive core properties of high breeding ratio, small inventory, compact size, low sodium void reactivity, and high transmutation ratio of the minor actinides. Among attractive cores, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void reactivity of less than 8${$}$, a core height of less than 150 cm, a maximum cladding temperature of 650 $$^circ$$C, and a fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 without blanket fuels.

JAEA Reports

Conceptual core design of the breeder type FLWR using recovered uranium as blanket fuel

Uchikawa, Sadao; Nakano, Yoshihiro; Okubo, Tsutomu; Kobayashi, Noboru

JAEA-Research 2008-040, 24 Pages, 2008/03

JAEA-Research-2008-040.pdf:8.5MB

The present study deals with the potential effects of utilization of uranium recovered from LWR spent fuels as the axial blanket fuels in the core of the breeder type FLWR, which is characterized by the five-layer double-flat-core structure. Design studies shows that the adoption of recovered UO$$_{2}$$ blanket can reduce the required plutonium inventory to attain the same core performances as the previous core design of depleted uranium blankets. Furthermore, the utilization of re-enriched recovered uranium with $$^{235}$$U enrichment of about 5wt% as the internal blanket fuel is evaluated to have a potential of a conversion ratio of 1.1. The core design with reduced plutonium inventory and high conversion ratio can contribute to realize a smooth and speedy transition from the LWR fuel cycle to the sustainable fuel cycle, as a core design for the transitioning period, during which a sufficient amount of uranium recovered from LWR spent fuels is available as resources.

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