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Kitazato, Kohei*; Milliken, R. E.*; Iwata, Takahiro*; Abe, Masanao*; Otake, Makiko*; Matsuura, Shuji*; Takagi, Yasuhiko*; Nakamura, Tomoki*; Hiroi, Takahiro*; Matsuoka, Moe*; et al.
Nature Astronomy (Internet), 5(3), p.246 - 250, 2021/03
Times Cited Count：44 Percentile：97.1(Astronomy & Astrophysics)Here we report observations of Ryugu's subsurface material by the Near-Infrared Spectrometer (NIRS3) on the Hayabusa2 spacecraft. Reflectance spectra of excavated material exhibit a hydroxyl (OH) absorption feature that is slightly stronger and peak-shifted compared with that observed for the surface, indicating that space weathering and/or radiative heating have caused subtle spectral changes in the uppermost surface. However, the strength and shape of the OH feature still suggests that the subsurface material experienced heating above 300 
C, similar to the surface. In contrast, thermophysical modeling indicates that radiative heating does not increase the temperature above 200 C at the estimated excavation depth of 1 m, even if the semimajor axis is reduced to 0.344 au. This supports the hypothesis that primary thermal alteration occurred due to radiogenic and/or impact heating on Ryugu's parent body.
Zr determination with isotope dilution ICP-MSAsai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*
Talanta, 185, p.98 - 105, 2018/08
Times Cited Count：8 Percentile：31.52(Chemistry, Analytical)Estimating the risks associated with radiation from long-lived fission products (LLFP) in radioactive waste is essential to ensure the long-term safety of potential disposal sites. In this study, the amount of Zr, a LLFP, was determined by ICP-MS after separating Zr from a spent nuclear fuel solution using a microvolume anion-exchange cartridge (TEDA cartridge). The TEDA cartridge achieved highly selective separation of Zr regardless of its small bed volume of 0.08 cm
. The time taken to complete the Zr separation was 1.2 min with a flow rate of 1.5 mL/min, which was 10 times faster than that for a conventional anion-exchange resin column. Almost all the other elements were removed, leading to accurate measurement of Zr. The result connects experimental value to theoretical prediction provided by ORIGEN2, which requires verification. With the measured value, we demonstrated that the theoretical value is reliable enough to estimate radiation risks.
Np content in spent nuclear fuelAsai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*
Analytical Chemistry, 88(6), p.3149 - 3155, 2016/03
Times Cited Count：8 Percentile：29.59(Chemistry, Analytical)Neptunium-237 (Np) is one of the major long-lived radionuclides found in spent nuclear fuel. To evaluate the long-term safety of a HLW repository, the Np content in spent nuclear fuel must be determined. In this study, micro-volume anion-exchange porous polymer disk-packed cartridges were prepared for Am-Np separation, which is required prior to the measurement of Np with ICP-MS. Disks with a volume of 0.08 cm were cut out from porous sheets having triethylenediamine (TEDA)-containing polymer chains densely attached on the pore surface. The resulting TEDA-introduced disk cartridge was applied to a spent nuclear fuel sample. The chemical yield of Np was 90.4%, which is sufficiently high for ICP-MS measurement of Np. Compared with the conventional separation technique using commercially available anion-exchange resin columns, the time required to adsorb, wash and elute Np using the TEDA-introduced disk cartridge was reduced by 75%.
Sakai, Takaaki; Yamano, Hidemasa; Tanaka, Masaaki; Ono, Ayako; Ohshima, Hiroyuki; Kaneko, Tetsuya*; Hirota, Kazuo*; Sago, Hiromi*; Xu, Y.*; Iwamoto, Yukiharu*; et al.
Proceedings of 15th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-15) (USB Flash Drive), 13 Pages, 2013/05
The development of flow-induced vibration evaluation methodology has reached a milestone that separate-effect experimental data under a high Reynolds number regime including swirl and deflected inflow conditions are available for the validation of the methodology. On the other hand, technical standards are desirable to be documented for designers of sodium-cooled fast reactors. From such a background, the documentation of a flow-induced vibration design guideline has been made for the hot-leg piping of Japan sodium-cooled fast reactor. This paper describes the design guideline of the flow-induced vibration evaluation methodology, which has been informed from main separate-effect experiments, as well as supplemental interpretation for the guideline.
Mo-
Tc domestic production with high-density MoO
pellets by (n,
) reactionTsuchiya, Kunihiko; Tanase, Masakazu*; Takeuchi, Nobuhiro*; Kobayashi, Masaaki*; Hasegawa, Yoshio*; Yoshinaga, Hideo*; Kaminaga, Masanori; Ishihara, Masahiro; Kawamura, Hiroshi
Proceedings of 5th International Symposium on Material Testing Reactors (ISMTR-5) (Internet), 10 Pages, 2012/10
As one of effective uses of the JMTR, JAEA has a plan to produce Mo by (n, ) method, a parent nuclide of Tc. In case of Japan, the supplying of Mo depends only on imports from foreign countries. The R&D on production method of Mo -Tc has been performed with Japanese industrial users under the cooperation programs. The main R&D items for the production are (1) Fabrication of irradiation target such as the sintered MoO pellets, (2) Separation and concentration of Tc by the solvent extraction from Mo solution, (3) Examination of Tc solution for a medicine, and (4) Mo recycling from Mo generator and solution. In this paper, the status of the R&D is introduced for the production of Mo -Tc.
Tochio, Daisuke; Hamamoto, Shimpei; Inoi, Hiroyuki; Shimazaki, Yosuke; Sekita, Kenji; Kondo, Masaaki; Saikusa, Akio; Kameyama, Yasuhiko; Saito, Kenji; Emori, Koichi; et al.
JAEA-Technology 2010-038, 57 Pages, 2010/12
JAEA-Technology-2010-038.pdf:2.36MB
In HTTR, in-service operation is conducted through the rise-to power operation with rated operation or high-temperature test operation from achievement of first criticality at 1998. To make practical use HTGR system, it must be demonstrated to supply stable heat to heat utilization system for long-term. In HTTR, high-temperature/parallel-loaded long-term operation had been performed from January 2010. As the result, it was demonstrated to supply stable heat to heat utilization system for 50 days with HTTR, moreover, various long-term operation data were gained. This paper reports the characteristics of the high-temperature long-term operation for HTTR obtained from the operation.
Maekawa, Yasunari; Kato, Jun*; Katakai, Masashi*; Ishihara, Masaaki*; Enomoto, Kazuyuki; Hagiwara, Tokio*; Ishii, Tatsuhito*; Ito, Kazuo*; Koshikawa, Hiroshi; Yoshida, Masaru
Macromolecular Chemistry and Physics, 209(6), p.625 - 633, 2008/03
Times Cited Count：2 Percentile：7.31(Polymer Science)A polysulfonamide and copolymers consisting of sulfonamides and amides underwent Fries rearrangement and scission, forming amino groups upon EB irradiation. The EB irradiation of these films with a dose of 500 
C/cm
followed by the color forming reaction provided the color imaging of line/space patterns with resolution of 300 nm.
Ohashi, Kazutaka; Nishihara, Tetsuo; Kunitomi, Kazuhiko; Nakano, Masaaki*; Tazawa, Yujiro*; Okamoto, Futoshi*
Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(1), p.32 - 43, 2008/03
Interests on the development of the Very High-Temperature Gas-Cooled Reactor (VHTR), of which the reactor outlet temperature is 950C or much higher, are recently increasing world-widely and it was selected as one of the candidate reactor types of the GIF. Japan Atomic Energy Agency has already initiated R&D efforts on the electricity and hydrogen co-generation plant with VHTR system, GTHTR300C. Although technical feasibility of its VHTR reactor using Pin-in-block fuel, which has experience to be already used in the HTTR, has been shown fundamentally, more improvements of the core performances, such as decrease of the occupational exposure doses during the plant maintenance, are desired. This report presents the results of the conceptual core design study using Multi-hole type fuel and the study on the occupational exposure doses. The latter results shows much better plant maintainability compared to the previous results of the GTHTR-300.
Higuchi, Hidekazu; Osugi, Takeshi; Nakashio, Nobuyuki; Momma, Toshiyuki; Tohei, Toshio; Ishikawa, Joji; Iseda, Hirokatsu; Mitsuda, Motoyuki; Ishihara, Keisuke; Sudo, Tomoyuki; et al.
JAEA-Technology 2007-038, 189 Pages, 2007/07
JAEA-Technology-2007-038-01.pdf:15.13MBJAEA-Technology-2007-038-02.pdf:38.95MBJAEA-Technology-2007-038-03.pdf:48.42MBJAEA-Technology-2007-038-04.pdf:20.53MBJAEA-Technology-2007-038-05.pdf:10.44MB
The Advanced Volume Reduction Facilities (AVRF) is constructed to manufacture the waste packages of radioactive waste for disposal in the Nuclear Science Research Institute of the Japan Atomic Energy Agency. The AVRF is constituted from two facilities. The one is the Waste Size Reduction and Storage Facility (WSRSF) which is for reducing waste size, sorting into each material and storing the waste package. The other is the Waste Volume Reduction Facility (WVRF) which is for manufacturing the waste package by volume reducing treatment and stabilizing treatment. WVRF has an induction melting furnace, a plasma melting furnace, an incinerator, and a super compactor for treatment. In this report, we summarized about the basic concept of constructing AVRF, the constitution of facilities, the specifications of machineries and the state of trial operation until March of 2006.
Sumita, Junya; Nakano, Masaaki*; Tsuji, Nobumasa*; Shibata, Taiju; Ishihara, Masahiro
JAERI-Tech 2004-055, 25 Pages, 2004/08
JAERI-Tech-2004-055.pdf:4.25MB
Neutron irradiation remarkably reduces the thermal conductivity of graphite, and the reduced thermal conductivity is recovered by annealing effect if the graphite is heated above the irradiation temperature. Therefore, it is expected that the reduced thermal conductivity of graphite components in the HTGR could be recovered by the annealing effect in accidents, such as a depressurization accident. Then, an analytical investigation of the annealing effect on thermal performance of a HTGR core was carried. The analysis showed that the annealing effect reduces the maximum fuel temperature about 70C, and it is important to introduce the annealing effect appropriately in the temperature analysis of the core components and reactor internals. In addition, an annealing test method was investigated to evaluate the effect quantitatively, and the test plan was made.
Np in spent nuclear fuelAsai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*
no journal, ,
no abstracts in English
Np in spent nuclear fuel using anion-exchange porous polymeric filterAsai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*
no journal, ,
no abstracts in English
Asai, Shiho; Hanzawa, Yukiko; Konda, Miki; Suzuki, Daisuke; Magara, Masaaki; Kimura, Takaumi; Ishihara, Ryo*; Saito, Kyoichi*; Yamada, Shinsuke*; Hirota, Hideyuki*
no journal, ,
Zr is a long-lived fission product which can be found in spent nuclear fuel and HLW. The estimation of theZr content is indispensable to achieve a safety disposal of HLW because Zr is predicted to be one of the major contributors to radiation dose. However, only a few measured Zr values have been reported, leading to a high demand for development of an efficient analytical method. Our group has been prepared a new porous filter cartridge which has densely bound ion-exchanger onto the pore surface of the filter, enabling a high-capacity and rapid adsorption. In order to apply this filter cartridge to a pretreatment for the Zr measurement with ICP-MS, an elution profiles of Zr and the other coexisting elements were examined. According to the resultant separation conditions, Zr in a spent nuclear fuel sample was successfully separated. The measured Zr content is 98.2 
5.1 ng, which agrees with the theoretical value.
