PSTEP: Project for solar-terrestrial environment prediction

Kusano, Kanya*; Ichimoto, Kiyoshi*; Ishii, Mamoru*; Miyoshi, Yoshizumi*; Yoden, Shigeo*; Akiyoshi, Hideharu*; Asai, Ayumi*; Ebihara, Yusuke*; Fujiwara, Hitoshi*; Goto, Tadanori*; et al.

Earth, Planets and Space (Internet), 73(1), p.159_1 - 159_29, 2021/12

https://doi.org/10.1186/s40623-021-01486-1

The PSTEP is a nationwide research collaboration in Japan and was conducted from April 2015 to March 2020, supported by a Grant-in-Aid for Scientific Research on Innovative Areas from the Ministry of Education, Culture, Sports, Science and Technology of Japan. It has made a significant progress in space weather research and operational forecasts, publishing over 500 refereed journal papers and organizing four international symposiums, various workshops and seminars, and summer school for graduate students at Rikubetsu in 2017. This paper is a summary report of the PSTEP and describes the major research achievements it produced.

Present status of J-PARC linac

Oguri, Hidetomo; Hasegawa, Kazuo; Ito, Takashi; Chishiro, Etsuji; Hirano, Koichiro; Morishita, Takatoshi; Shinozaki, Shinichi; Ao, Hiroyuki; Okoshi, Kiyonori; Kondo, Yasuhiro; et al.

Proceedings of 11th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.389 - 393, 2014/10

http://www.pasj.jp/web_publish/pasj2014/proceedings/PDF/FSP0/FSP021.pdf

no abstracts in English

Corrosion behavior of Fe, Cr, Ni and their alloys in sodium molybdate melts

Nagai, Takayuki; Kikuchi, Kotaro*; Kano, Yoshiharu*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 12(1), p.76 - 88, 2013/03

https://doi.org/10.3327/taesj.J12.013

To understand the corrosion behavior of stainless steel and Ni-Cr alloys in sodium molybdate melts, the immersed corrosion tests of Fe, Cr, Ni, SUS316, 55Ni-45Cr, and Inconel 690 specimens were carried out in NaMoO and NaMoO melts at 750 C. The results of the immersed test, it was found that the corrosion rate decreased with an increase of chromium content in chemical composition of specimens. It was observed that CrO formed on the surface of Cr, 55Ni-45Cr, and Inconel 690 specimens by immersed in NaMoO melt, which corrosion rates were low in the melt, and we presumed that this CrO on the surface controlled the corrosion behavior of their material in the melt.

Development of separation technology of transuranium elements and fission products by using new extractants and adsorbents; Development of separation technology of Cs and Sr (Contract research)

Hoshi, Harutaka; Kikuchi, Takahiro; Asakura, Toshihide; Morita, Yasuji; Kimura, Takaumi

JAEA-Research 2010-016, 70 Pages, 2010/07

JAEA-Research-2010-016.pdf:2.31MB

We have studied selective separation of Cs and Sr, which are included in high level liquid waste (HLLW) generated from reprocessing of spent nuclear fuel and are major heat generators, by using extractant impregnated adsorbents. Cs adsorbent using calix arane derivatives showed excellent selectivity for Cs. It also showed significant stability against -irradiation. Sr adsorbent using crown ether derivatives also showed high selectivity for Sr from nitric acid solution, except for Ba and Tc. Dynamic capacity decreased ca. 30% after -irradiation. Hot test using genuine HLLW stored in NUCEF was performed for separation of Cs and Sr through columns, respectively. Each Cs and Sr was separated from other typical fission product elements as well as the results obtained in preliminary experiments. Finally, Cs and Sr were separated according to a supposed separation scheme. Although some complexing agents were added in simulate HLLW, no negative effect was found.

Study on stability of CsSr solvent impregnated resin against irradiation

Hoshi, Harutaka; Kikuchi, Takahiro; Morita, Yasuji; Kimura, Takaumi

JAEA-Review 2009-041, JAEA Takasaki Annual Report 2008, P. 24, 2009/12

https://jopss.jaea.go.jp/pdfdata/JAEA-Review-2009-041.pdf

Minimization of radioactive waste from reprocessing process of spent nuclear fuel is strongly desired. We develop an advanced technology for separation of heat generating elements (Cs and Sr) from high level waste to optimize radioactive waste by its characteristics. Some novel solvent impregnated resins (SIRs) were prepared and these SIRs indicated promising ability to separate Cs and Sr from other typical fission products dissolved in nitric acid solution. These adsorbents contacting with nitric acid solution were exposed to ray in a vial. After the irradiation, adsorbents were filtrated through a membrane filter. The adsorption capacity was examined by using irradiated adsorbents after drying. It is concluded that both SIRs maintained their high selectivity for Cs and Sr, respectively, after irradiation. Decreasing ratio of adsorption capacity by irradiation can be estimated from these results. It contributes significantly for conceptual design of separation plant.

Corrosion behavior of SUS316 clad in alkalis molybdate melt

Nagai, Takayuki; Kikuchi, Kotaro*; Kano, Yoshiharu*; Fukushima, Mineo

Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(4), p.370 - 379, 2008/12

https://doi.org/10.3327/taesj.J07.050

Study of a new pyrochemical processing using alkaline molybdate melt has been carrying out as a candidate reprocessing process for spent oxide fuels. In our previous study, we had confirmed that melting of UO pellet into a molten NaMoO-MoO and electrolytic recovering of UO from the melt included uranyl ions. In this report, the corrosion behavior of the SUS316 cladding tube in molten NaMoO-MoO was evaluated, and we confirmed the dissolution of SUS316 elements into the melt when the amount of additional MoO was excess. And, it is necessary to understand an appropriate amount of additional MoO since the result of this study.

Imaging of the Mozumi-Sukenobe fault, Hida district, central Japan, by the seismic reflection method

Ito, Tanio*; Tsumura, Noriko*; Takeuchi, Akira*; Ishimaru, Tsuneari; Takami, Akira*; Ikawa, Hidemasa*; Komada, Nozomi*; Yamamoto, Shuji*; Kikuchi, Shinsuke*; Miyauchi, Takahiro*; et al.

Geodynamics of Atotsugawa Fault System, p.17 - 24, 2007/00

no abstracts in English

Advanced fusion technologies developed for JT-60 superconducting Tokamak

Sakasai, Akira; Ishida, Shinichi; Matsukawa, Makoto; Akino, Noboru; Ando, Toshinari*; Arai, Takashi; Ezato, Koichiro; Hamada, Kazuya; Ichige, Hisashi; Isono, Takaaki; et al.

Nuclear Fusion, 44(2), p.329 - 334, 2004/02

https://doi.org/10.1088/0029-5515/44/2/015

no abstracts in English

Design of a superconducting coil system for remodeling JT-60

Ando, Toshinari*; Ishida, Shinichi; Kato, Takashi; Kikuchi, Mitsuru; Kizu, Kaname; Matsukawa, Makoto; Miura, Yushi; Nakajima, Hideo; Sakasai, Akira; Sugimoto, Makoto; et al.

IEEE Transactions on Applied Superconductivity, 12(1), p.500 - 503, 2002/03

https://doi.org/10.1109/TASC.2002.1018452

no abstracts in English

ATR Demonstration reactor integrity verification test for the pressur tube rolled joint portion (Fiscal 1988)

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PNC TN9410 94-052, 251 Pages, 1994/01

PNC-TN9410-94-052.pdf:9.33MB

The structure of the pressure tube rolled joint portion for the ATR Demonstration Reactor is somewhat changed from that for Fugen, in order reduce the residual stress around the portion. Therefore, Constant Temperature Endurance Test and Thermal Cycle Endurance Test have been conducted under the reactor operating conditions except irradiation to examine the rolled joint integrity. (1)Constant Temperature Endurance Test. In fiscal 1988, Constant Temperature Endurance Test have been performed in the Component Test Loop for 2,033 hours under the reactor operating conditions (pressure 75kg/cm, temperature 280C) for the JP-3 specimen (total testing period 4,033 hours) and the JP-4, JP-5 specimens (total testing period 9,533 hours). After the endurance test it was found by the helium leak test that the rolled joint tightness was maintained enough. Therefore, it was confirmed that the reduction of the residual atress at the rolled joint portion which occurred at the initial stage of the operation did not affect the rolled joint tightness. (2)Thermal Cycle Endurance Test. The helium leak test and the ultrasonic flaw detection test were performed before and after Thermal Cycle Endurance Test (the unmber of thermal cycle increaced from 60 times to 140 times in fisical 1988), of which results showed the integrity of the rolled joint tightness and no crack propagation near the rolled joint portion. Therefore, it was confirmed that 140 times thermal cycle which was the design value for the reactor life of 30 years did not affect the rolled joint tightness and the initiation and propagation of a crack even for the specimen with 200 ppm hydrogen.