Annual report for FY2021 on the activities of Naraha Center for Remote Control Technology Development (April 1, 2021 - March 31, 2022)

Akiyama, Yoichi; Shibanuma, So; Yanagisawa, Kenichi*; Yamada, Taichi; Suzuki, Kenta; Yoshida, Moeka; Ono, Takahiro; Kawabata, Kuniaki; Watanabe, Kaho; Morimoto, Kyoichi; et al.

JAEA-Review 2023-015, 60 Pages, 2023/09

JAEA-Review-2023-015.pdf:4.78MB

Naraha Center for Remote Control Technology Development (NARREC) was established in Japan Atomic Energy Agency to promote a decommissioning work of Fukushima Daiichi Nuclear Power Station (Fukushima Daiichi NPS). NARREC consists of a Full-scale Mock-up Test Building and Research Management Building. Various test facilities are installed in these buildings for the decommissioning work of Fukushima Daiichi NPS. These test facilities are intended to be used for various users, such as companies engaged in the decommissioning work, research and development institutions, educational institutions and so on. The number of NARREC facility uses was 84 in FY2021. We participated booth exhibitions and presentations on the decommissioning related events. Moreover, we also contributed to the development of human resources by supporting the 6th Creative Robot Contest for Decommissioning. As a new project, "Narahakko Children's Classroom" was implemented for elementary school students in Naraha Town. This report summarizes the activities of NARREC in FY2021, such as the utilization of facilities and equipment of NARREC, the development of remote-control technologies for supporting the decommissioning work, arrangement of the remote-control machines for emergency response, and training for operators by using the machines.

Establishing an evaluation method for the aging phenomenon by physical force in fuel debris

Suzuki, Seiya; Arai, Yoichi; Okamura, Nobuo; Watanabe, Masayuki

Journal of Nuclear Science and Technology, 60(7), p.839 - 848, 2023/07

https://doi.org/10.1080/00223131.2022.2159559

The fuel debris, consisting of nuclear fuel materials and reactor structural materials, generated in the accident of Fukushima Daiichi Nuclear Power Plant can become deteriorated like rocks under the changes of environmental temperature. Although the fuel debris have been cooled by water for 10 years, they are affected by seasonal and/or day-and-night temperature changes. Therefore, in evaluating the aging behavior of the fuel debris, it is essential to consider the changes in environmental temperature. Assuming that the fuel debris are deteriorated, radioactive substances that have recently undergone micronization could be eluted into the cooling water, and such condition may affect defueling methods. We focused on the effect of repeated changes in environmental temperature on the occurrence of cracks, and an accelerated test using simulated fuel debris was carried out. The length of the crack increases with increasing number of heat cycle; therefore, the fuel debris become brittle by stress caused by thermal expansion and contraction. In conclusion, it was confirmed that the mechanical deterioration of the fuel debris is similar to that of rocks or minerals, and it became possible to predict changes in the length of the crack in the simulated fuel debris and environmental model.

Investigation of adsorption mechanism of Mo(VI) by baker's yeast and applicability to the uranium liquid waste treatment process

Arai, Yoichi; Hasegawa, Kenta; Watanabe, So; Watanabe, Masayuki; Minowa, Kazuki*; Matsuura, Haruaki*; Hagura, Naoto*; Katsuki, Kenta*; Arai, Tsuyoshi*; Konishi, Yasuhiro*

Journal of Radioanalytical and Nuclear Chemistry, 9 Pages, 2023/00

https://doi.org/10.1007/s10967-023-09278-5

Clogging properties of HEPA filter induced by loading of soot from burned glove-box panel materials

Tashiro, Shinsuke; Ono, Takuya; Amano, Yuki; Yoshida, Ryoichiro; Watanabe, Koji*; Abe, Hitoshi

Nuclear Technology, 208(10), p.1553 - 1561, 2022/10

https://doi.org/10.1080/00295450.2022.2045179

To contribute to the confinement safety evaluation of the radioactive materials in the Glove box (GB) fire accident, combustion tests with the Polymethyl methacrylate (PMMA) and the Polycarbonate (PC) as typical panel materials for the GB have been conducted with a relatively large scale apparatus. As the important data for evaluating confinement safety, the release ratio and the particle size distribution of the soot generated from the burned materials were obtained. Furthermore, the rise of the differential pressure (P) of the high efficiency particle air (HEPA) filter by the soot loading was also investigated. As results, the release ratio of the soot from the PC was about seven times as large as the PMMA. In addition, it was found that the behavior of the rise of the P with soot loading could be represented uniformly regardless of kinds of combustion materials by considering effect of the loading volume of the soot particle in the relatively low loading region.

Measurements of thermal conductivity for near stoichiometric (UPuAm)O (z = 0.05, 0.10, and 0.15)

Yokoyama, Keisuke; Watanabe, Masashi; Tokoro, Daishiro*; Sugimoto, Masatoshi*; Morimoto, Kyoichi; Kato, Masato; Hino, Tetsushi*

Nuclear Materials and Energy (Internet), 31, p.101156_1 - 101156_7, 2022/06

https://doi.org/10.1016/j.nme.2022.101156

In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this study, The thermal conductivities of near stoichiometric (UPuAm)O solid solutions(z = 0.05, 0.10, and 0.15) have been measured between room temperature (RT) and 1473 K. The thermal conductivities decreased with increasing Am content and satisfied the classical phonon transport model ((A+BT)) up to about 1473 K. A values increased linearly with increasing Am content because the change in ionic radius affects the conduction of the phonon due to the solid solution in U and Am. B values were independent of Am content.

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.

Comprehensive treatment of radioactive liquid waste of Chemical Processing Facility

Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Watanabe, So; Shibata, Atsuhiro; Nomura, Kazunori

JAEA-Technology 2021-007, 27 Pages, 2021/06

JAEA-Technology-2021-007.pdf:2.43MB

Chemical Processing Facility (CPF) of Japan Atomic Energy Agency (JAEA) has been developing the fast reactor fuel reprocessing and vitrification technology. The various kinds of radioactive liquid wastes, which were generated by those experiments and analysis, stored in the hot cells and glove boxes of CPF. The treatment of radioactive liquid wastes were started since July 2015; however, treatment of several kinds of liquid wastes are revealed to be difficult due to contain the various hazardous chemicals. Therefore, in order to establish the new technology suitable for radioactive liquid waste treatment, several collaborative research programs with several universities and national research organizations were started. The combined project lead by JAEA was named to be STRAD (Systematic Treatments of Radioactive liquid wastes for Decommissioning) project. In this project, the process flow for treatment of several actual liquid wastes were established. In this report, treated method and progress of actual liquid wastes of CPF are summarized.

Physical properties of non-stoichiometric (U,Pu)O

Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Morimoto, Kyoichi; Kato, Masato

2018 GIF Symposium Proceedings (Internet), p.315 - 320, 2020/05

Recently, a research group studying at Plutonium Fuel Development Facility (PFDF) in Japan Atomic Energy Agency has systematically measured vast amounts of physical properties in the non-stoichiometric (U, Pu)O. Lattice parameter, elastic modulus, thermal expansion, oxygen potential, oxygen chemical diffusion coefficient and thermal conductivity were successfully measured as function of Pu content, O/M ratio and temperature, and the effects of Pu content and O/M ratio on their physical properties were evaluated. In this work, those experimental data are reviewed, and latest experimental data set on the non-stoichiometric (U, Pu)O are presented. The data set would be available in development of a fuel performance code.

STRAD project for systematic treatments of radioactive liquid wastes generated in nuclear facilities

Watanabe, So; Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Takahatake, Yoko; Shibata, Atsuhiro; Nomura, Kazunori; Kamiya, Yuichi*; Asanuma, Noriko*; Matsuura, Haruaki*; et al.

Progress in Nuclear Energy, 117, p.103090_1 - 103090_8, 2019/11

AA2019-0193.pdf:1.29MB

https://doi.org/10.1016/j.pnucene.2019.103090

Analysis on adsorbent for spent solvent treatment by micro-PIXE and EXAFS

Arai, Yoichi; Watanabe, So; Ono, Shimpei; Nakamura, Masahiro; Shibata, Atsuhiro; Nakamura, Fumiya*; Arai, Tsuyoshi*; Seko, Noriaki*; Hoshina, Hiroyuki*; Hagura, Naoto*; et al.

International Journal of PIXE, 29(1&2), p.17 - 31, 2019/00

https://doi.org/10.1142/S0129083519500128

The spent PUREX solvent containing U and Pu is generated from the reprocessing process of spent nuclear fuel. The nuclear material removal is important for the safe storage or disposal of the spent solvent. Our previous study revealed that the adsorbent with the iminodiacetic acid (IDA) functional group is one of the most promising materials for designing the nuclear material recovery process. Accordingly, an IDA-type adsorbent was synthesized by using graft polymerization technology or a chemical reaction to improve the adsorption rate and capacity. The synthesized IDA-type adsorbent was characterized by micro particle-induced X-ray emission (PIXE) and extended X-ray absorption fine structure (EXAFS) analyses. The micro-PIXE analysis revealed that Zr was adsorbed on the whole synthesized adsorbents and quantified the microamount of adsorbed Zr. Moreover, EXAFS suggested that Zr in the aqueous solution and solvent can be trapped by the IDA group with different mechanisms.