Current status of JAEA-AMS-TONO in the 20th year

Kokubu, Yoko; Fujita, Natsuko; Miyake, Masayasu; Watanabe, Takahiro; Ishizaka, Chika; Okabe, Nobuaki; Ishimaru, Tsuneari; Matsubara, Akihiro*; Nishizawa, Akimitsu*; Nishio, Tomohiro*; et al.

Nuclear Instruments and Methods in Physics Research B, 456, p.271 - 275, 2019/10

https://doi.org/10.1016/j.nimb.2019.03.015

JAEA-AMS-TONO has been in operation at the Tono Geoscience Center, Japan Atomic Energy Agency since 1998 and 20 years have passed from the beginning of its utilization. The AMS system is a versatile system based on a 5 MV tandem Pelletron type accelerator. The system has been used to measure carbon-14 (C), beryllium-10 (Be) and aluminium-26 (Al). In addition, the development of measurement of iodine-129 (I) has been started. The main use is measurement of C in geological samples for dating studies in neotectonics and hydrogeology. In order to increase the speed of sample preparation, we introduced the automated graphitization equipment and made a gas-strip line to collect dissolved inorganic carbon in groundwater samples. Measurement of Be and Al has been used for geoscience studies and the detection limit in the measurement of Be was improved by Be-counting suppression. Recently tuning of measurement condition of I has been progressed.

The present status of the AGE3 automated graphitization equipment for radiocarbon dating at the JAEA-AMS-TONO

Watanabe, Takahiro; Kokubu, Yoko; Fujita, Natsuko; Ishizaka, Chika*; Nishio, Tomohiro; Matsubara, Akihiro*; Miyake, Masayasu; Kato, Motohisa*; Isozaki, Nobuhiro*; Torazawa, Hitoshi*; et al.

JAEA-Conf 2018-002, p.116 - 119, 2019/02

https://doi.org/10.11484/jaea-conf-2018-002

AMS is widely used for radiocarbon dating of geological samples. However, improvement in efficiency of sample preparation techniques are needed for high-time resolution dataset. In 2016, automated graphitization equipment (AGE3, IonPlus AG) has been installed in Toki Research Institute of Isotope Geology and Geochronology, Tono Geoscience Center, JAEA. Background values and carbon recovery rates during preparation process of AGE3 should be estimated before application in radiocarbon dating. In this study, the AGE3 system was evaluated using the international standard materials (IAEA-C1, C4, C5, C6, C7, C9 and NIST-SRM4990C) at JAEA-AMS-TONO. Graphite samples was prepared by the AGE3 system and radiocarbon concentration of these standards was measured by AMS. The results were agreement with the consensus values. Background values were 0.150.01 pMC (IAEA-C1) using the AGE3 system. Therefore, we concluded that the system can be adapted for radiocarbon dating of geological samples.

Present status of the accelerator facility at the JAEA-AMS-TONO; 2016

Fujita, Natsuko; Miyake, Masayasu; Watanabe, Takahiro; Kokubu, Yoko; Ishimaru, Tsuneari; Matsubara, Akihiro*; Nishio, Tomohiro*; Kato, Motohisa*; Isozaki, Nobuhiro*; Torazawa, Hitoshi*; et al.

JAEA-Conf 2018-013, p.96 - 99, 2019/02

https://doi.org/10.11484/jaea-conf-2018-003

no abstracts in English

Present status of the JAEA-AMS-TONO; 2016

Fujita, Natsuko; Miyake, Masayasu; Watanabe, Takahiro; Kokubu, Yoko; Ishimaru, Tsuneari; Matsubara, Akihiro*; Isozaki, Nobuhiro*; Nishio, Tomohiro*; Kato, Motohisa*; Torazawa, Hitoshi*; et al.

Dai-19-Kai AMS Shimpojiumu, 2016-Nendo "Jumoku Nenrin" Kenkyukai Kyodo Kaisai Shimpojiumu Hokokushu, p.68 - 71, 2017/06

no abstracts in English

Enhancement of the methodology of repository design and post-closure performance assessment for preliminary investigation stage, 3; Progress report on NUMO-JAEA collaborative research in FY2013 (Joint research)

Shibata, Masahiro; Sawada, Atsushi; Tachi, Yukio; Makino, Hitoshi; Wakasugi, Keiichiro; Mitsui, Seiichiro; Kitamura, Akira; Yoshikawa, Hideki; Oda, Chie; Ishidera, Takamitsu; et al.

JAEA-Research 2014-030, 457 Pages, 2015/03

JAEA-Research-2014-030.pdf:199.23MB

JAEA and NUMO have conducted a collaborative research work which is designed to enhance the methodology of repository design and post-closure performance assessment in preliminary investigation stage. With regard to (1) study on rock suitability in terms of hydrology, based on some examples of developing method of hydro-geological structure model, acquired knowledge are arranged using the tree diagram, and model uncertainty and its influence on the evaluation items were discussed. With regard to (2) study on scenario development, the developed approach for "defining conditions" has been reevaluated and improved from practical viewpoints. In addition, the uncertainty evaluation for the effect of use of cementitious material, as well as glass dissolution model, was conducted with analytical evaluation. With regard to (3) study on setting radionuclide migration parameters, based on survey of precedent procedures, multiple-approach for distribution coefficient of rocks was established, and the adequacy of the approach was confirmed though its application to sedimentary rock and granitic rock. Besides, an approach for solubility setting was developed including the procedure of selection of solubility limiting solid phase. The adequacy of the approach was confirmed though its application to key radionuclides.

Analysis on temperature distribution of reactor vessel cooling system during loss of coolant flow in HTTR

Takeda, Tetsuaki*; Ichimiya, Koichi*; Nishio, Hitoshi*; Nakagawa, Shigeaki; Takamatsu, Kuniyoshi

Proceedings of 7th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-7) (CD-ROM), 11 Pages, 2008/10

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The partial loss of coolant flow test was performed under the condition of the ATWS (Anticipated Transient without Scram). We are planning to perform the test of loss of coolant flow and stopping the vessel cooling system (VCS). The test of the loss of coolant flow as one of safety demonstration tests is carried out by tripping all gas circulators, and the stopping VCS test is performed continuously after the loss of coolant flow. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the VCS during the tests. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.

Removal of electric-charged aerosols by radiation effect

Nishio, Gunji; Takada, Junichi; ; Murata, Mikio; Abe, Hitoshi;

JAERI-Research 95-064, 45 Pages, 1995/09

JAERI-Research-95-064.pdf:1.61MB

no abstracts in English

Development of the computer code TRANS-ACE for analysing radioactive materials transport in reporocessing plants under accidental conditions

Nishio, Gunji; Abe, Hitoshi; Naito, Yoshitaka

JAERI-M 89-014, 169 Pages, 1989/02

JAERI-M-89-014.pdf:3.48MB

no abstracts in English

Effect of thermal radiation on temperature distribution of reactor pressure vessel in HTGR

Takeda, Tetsuaki; Nishio, Hitoshi*; Ichimiya, Koichi*

no journal, , 

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. It is necessary to consider the effect of thermal radiation from the RPV for evaluation of temperature of the VCS and concrete vessel.

Thermal hydraulic analysis on reactor pressure vessel in HTGR

Takeda, Tetsuaki; Nishio, Hitoshi*; Ichimiya, Koichi*

no journal, , 

Safety demonstration tests using the HTTR (High Temperature Engineering Test Reactor) are being performed to verify the inherent safety features and to validate the numerical code for the safety assessment of the VHTR (Very High Temperature Reactor). The test of loss of coolant flow as one of safety demonstration tests is carried out by tripping of the helium gas circulators. The objective of this study is to evaluate the temperature distribution of the reactor pressure vessel (RPV) and the vessel cooling system (VCS) during the loss of coolant flow. The temperature distribution of the RPV and surrounding concrete structure were obtained using a commercially available analysis code STAR-CD. The effect of thermal radiation from the RPV was evaluated using the analytical and experimental results. It was found that the RPV and surrounding concrete structure were cooled enough by the VCS in the case of loss of coolant flow.