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Kawase, Keiichi
Kankyo Hoshano Josen Gakkai-Shi, 9(4), p.171 - 172, 2021/12
no abstracts in English
Fukaya, Yuji
Annals of Nuclear Energy, 144, p.107503_1 - 107503_7, 2020/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Disposal scenario of High Level Waste (HLW) of Light Water Reactor (LWR) have been optimized to reduce waste volume and repository footprint in geological disposal. The optimization was performed with existing and demonstrated technology in the situation where the reprocessing will be delayed. In general, the scenario with Partitioning and Transmutation (P&T) is optimized to minimize waste package number generated in the situation where the spent fuel will be reprocessed immediately with the minimum cooling time. With considering the delay of reprocessing, it is found that the more simplified and effective optimization with the high-waste-loading glass and cold crucible induction melter technologies and without partitioning. The optimized case can achieve significant reduction of number of waste package generation and the repository footprint to half of those of non-P&T case with 100 years cooling.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Okumura, Masahiko; Kerisit, S.*; Bourg, I. C.*; Lammers, L. N.*; Ikeda, Takashi*; Sassi, M.*; Rosso, K. M.*; Machida, Masahiko
Journal of Environmental Radioactivity, 189, p.135 - 145, 2018/09
Times Cited Count：51 Percentile：87.2(Environmental Sciences)no abstracts in English
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi; Nishihara, Tetsuo; Tsubata, Yasuhiro; Matsumura, Tatsuro
Annals of Nuclear Energy, 116, p.224 - 234, 2018/06
Times Cited Count：2 Percentile：20.93(Nuclear Science & Technology)Optimization of disposal method and scenario to reduce volume of High Level Waste (HLW) and the footprint in a geological repository for High Temperature Gas-cooled Reactor (HTGR) has been performed. It was found that HTGR has great advantages to reducing HLW volume and its footprint, which are high burn-up, high thermal efficiency and pin-in-block type fuel, compared with those of LWR and has potential to reduce those more in the previous study. In this study, the scenario is optimized, and the geological repository layout is designed with the horizontal emplacement based on the KBS-3H concept instead of the vertical emplacement based on KBS-3V concept employed in the previous study. As a result, for direct disposal, the repository footprint can be reduced by 20 % by employing the horizontal without change of the scenario. By extending 40 years for cooling time before disposal, the footprint can be reduced by 50 %. For disposal with reprocessing, the number of canister generation can be reduced by 20 % by extending cooling time of 1.5 years between the discharge and reprocessing. The footprint per electricity generation can be reduced by 80 % by extending 40 years before disposal. Moreover, by employing four-group partitioning technology without transmutation, the footprint can be reduced by 90 % with cooling time of 150 years.
Sakamoto, Hiroyuki*; Akagi, Yosuke*; Yamada, Kazuo*; Tachi, Yukio; Fukuda, Daisuke*; Ishimatsu, Koichi*; Matsuda, Mikiya*; Saito, Nozomi*; Uemura, Jitsuya*; Namihira, Takao*; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 17(2), p.57 - 66, 2018/05
Concrete debris contaminated with radioactive cesium and other nuclides have been generated from the accident in the Fukushima Dai-ichi Nuclear Power Plant and there will be generated due to the decommissioning of nuclear power plants in the future. Although conventional decontamination techniques are effective for flat concrete surfaces such as floors and walls, it is not clear what techniques to apply for decontaminating radioactive concrete debris. In this study, focusing on a pulsed power discharge technique, fundamental experimental works were carried out. Decontamination of concrete by applying the aggregate recycling technique using the pulsed power discharge technique was evaluated by measuring radioactivity of aggregate and sludge separated from the contaminated concrete. The results suggest that the separation into aggregate and sludge of the contaminated concrete debris could achieve decontamination and volume reduction of the radioactive concrete debris.
Sakauchi, Hitoshi; Kikuchi, Yuki; Imaizumi, Haruki; Fukui, Yasutaka
Dekomisshoningu Giho, (57), p.34 - 42, 2018/03
OWTF (Oarai Waste Reduction Treatment Facility) is constructed for volume reduction processing and stabilization treatment of radioactive solid waste, which generate from hot facilities in Oarai Research and Development Center of Japan Atomic Energy Agency, using in-can type high frequency induction heating by remote control. In this report, we describe the outline of OWTF under construction and treatment technologies, in which incinerating and melting.
Yaita, Tsuyoshi; Honda, Mitsunori; Shimoyama, Iwao; Ito, Kenichi*; Mampuku, Yuzo*; Tsuji, Takuya; Matsumura, Daiju
Nihon Genshiryoku Gakkai-Shi ATOMO
, 59(8), p.483 - 487, 2017/08
no abstracts in English
Kato, Jun; Nakagawa, Akinori; Taniguchi, Takumi; Sakakibara, Tetsuro; Nakazawa, Osamu; Meguro, Yoshihiro
JAEA-Review 2017-015, 173 Pages, 2017/07
JAEA-Review-2017-015.pdf:6.67MB
Various radioactive wastes have been generated at the Fukushima Daiichi Nuclear Power Station (1F). To dispose of the wastes underground, it is necessary to make a suitable waste package by the volume reduction and solidification of the wastes. To plan the future decommissioning of 1F, it is also necessary to estimate feasibility of existing treatment technology for those wastes. Therefore the document survey has been performed about volume reduction and solidification technologies that have domestic or foreign experiences of practical treatment for radioactive wastes to assist selection of suitable treatment of the wastes. This report shows the arranged results. The 1F wastes are classified into two groups, homogeneous particulate and liquid wastes and heterogeneous solid wastes. The needful items for the feasibility study such as a technology name, a fundamental principle, treatment efficiency, and characteristic of solidified waste are summarized in each group.
Ozawa, Takayuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
In a recycle system for minor actinides (MAs) currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. MA content is expected to be 
5 wt.% in the future recycle system, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important behaviors would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. The MA-containing influences were evaluated with CEPTAR.V2, including fuel properties and analysis models to evaluate the MA-MOX fuel irradiation behavior, by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo. The irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed the influences of Am-containing.
Ishihara, Keisuke; Yokota, Akira; Kanazawa, Shingo; Iketani, Shotaro; Sudo, Tomoyuki; Myodo, Masato; Irie, Hirobumi; Kato, Mitsugu; Iseda, Hirokatsu; Kishimoto, Katsumi; et al.
JAEA-Technology 2016-024, 108 Pages, 2016/12
JAEA-Technology-2016-024.pdf:29.74MB
Radioactive isotope, nuclear fuel material and radiation generators are utilized in research institutes, universities, hospitals, private enterprises, etc. As a result, various low-level radioactive wastes (hereinafter referred to as non-nuclear radioactive wastes) are produced. Disposal site for non-nuclear radioactive wastes have not been settled yet and those wastes are stored in storage facilities of each operator for a long period. The Advanced Volume Reduction Facilities (AVRF) are built to produce waste packages so that they satisfy requirements for shallow underground disposal. In the AVRF, low-level beta-gamma solid radioactive wastes produced in the Nuclear Science Research Institute are mainly treated. To produce waste packages meeting requirements for disposal safely and efficiently, it is necessary to cut large radioactive wastes into pieces of suitable size and segregate those depending on their types of material. This report summarizes activities of pretreatment to dispose of non-nuclear radioactive wastes in the AVRF.
Fukaya, Yuji; Nishihara, Tetsuo
Nuclear Engineering and Design, 307, p.188 - 196, 2016/10
Times Cited Count：4 Percentile：36.53(Nuclear Science & Technology)Reduction of High Level Waste (HLW) and footprint in a geological repository due to high burn-up and high thermal efficiency of High Temperature Gas-cooled Reactor (HTGR) has been investigated. A helium-cooled and graphite-moderated commercial HTGR was designed as a Gas Turbine High Temperature Reactor (GTHTR300), and the features are significantly high burn-up of approximately 120 GWd/t, high thermal efficiency around 50%, and pin-in-block type fuel. The pin-in-block type fuel was employed to reduce processed graphite volume in reprocessing, and effective waste loading method for direct disposal is proposed by applying the feature in this study. As a result, it is found that the number of canisters and its repository footprint per electricity generation can be reduced by 60% compared with LWR representative case for direct disposal because of the higher burn-up, higher thermal efficiency, less TRU generation, and effective waste loading proposed in this study for HTGR. For disposal with reprocessing, the number of canisters and its repository footprint per electricity generation can be reduced by 30% compared with LWR because of the 30% higher thermal efficiency of HTGR.
Sakauchi, Hitoshi; Sato, Isamu*; Donomae, Yasushi; Kitamura, Ryoichi
JAEA-Technology 2015-059, 352 Pages, 2016/03
JAEA-Technology-2015-059.pdf:51.53MB
OWTF (Oarai Waste Reduction Treatment Facility) is constructed for volume reduction processing and stabilization treatment of 
solid waste, which was generated from hot facilities in Oarai Research and Develop Center of Japan Atomic Energy Agency, using in-can type high frequency induction heating by remote control. This report describes corroborative tests, in which incinerating and melting performance for OWTF is confirmed with a full-scale testing furnace. We have been carrying out the tests of incinerating and melting treatment with some kinds of simulated wastes, such as enclosure form of radioactive wastes, material and articles.
Nakashio, Nobuyuki; Osugi, Takeshi; Iseda, Hirokatsu; Tohei, Toshio; Sudo, Tomoyuki; Ishikawa, Joji; Mitsuda, Motoyuki; Yokobori, Tomohiko; Kozawa, Kazushige; Momma, Toshiyuki; et al.
Journal of Nuclear Science and Technology, 53(1), p.139 - 145, 2016/01
Times Cited Count：1 Percentile：10.71(Nuclear Science & Technology)no abstracts in English
Sector of Fukushima Research and Development
JAEA-Evaluation 2015-004, 96 Pages, 2015/09
JAEA-Evaluation-2015-004.pdf:2.95MB
Japan Atomic Energy Agency (JAEA) consulted an assessment committee, "Evaluation Committee of Research and Development Activities for Fukushima Environment" (Committee) for ex-post and ex-ante evaluations of "Research and Development of the technologies for Fukushima Environment Recovery" in accordance with national guidelines and JAEA's regulations including "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by Cabinet Office. In response to the JAEA's request, the Committee evaluated the R&D activities during the 2nd Period of Mid-Term Research Plan (from March 2011 to March 2015) as ex-post evaluation and the R&D plans during the 3rd Period of Mid- to Long-Term Research Plan (from April 2015 to March 2022) as ex-ante evaluation. The Committee evaluated the R&D activities and programs based on the explanatory materials prepared by JAEA and its oral presentations with Q&A sessions. This report presents the results of the evaluation.
Yaita, Tsuyoshi; Kobayashi, Toru; Ikeda, Takashi; Matsumura, Daiju; Machida, Masahiko; Okumura, Masahiko; Nakamura, Hiroki
Hoshako, 27(6), p.315 - 322, 2014/11
no abstracts in English
Yaita, Tsuyoshi
Petorotekku, 37(5), p.329 - 333, 2014/05
no abstracts in English
Nakashio, Nobuyuki; Nakashima, Mikio
Dekomisshoningu Giho, (26), p.45 - 55, 2002/11
Melting treatment of low-level radioactive wastes (LLW) is considered to be a promising technology for the preparation of a stable solid that will be disposed of in near surface repositories. This is because of large reduction of waste volume and production of a stable homogeneous solidified product. In the Japan Atomic Energy Research Institute (JAERI), the construction of the Waste Volume Reduction Facilities (WVRF) has been in progress since 1999. In advance of operation of the WVRF, we have been conducting melting tests of non-metallic solid wastes with the aim of establishing the optimum melting condition for preparation of a stable solid that is suitable for disposal. We have reviewed a part of the melting test conducted in our program.
; Murata, Eiichi*; Sawahata, Yoshikazu*; Saito, Akira*
JNC TN8430 2001-002, 43 Pages, 2001/02
JNC-TN8430-2001-002.pdf:1.98MB
Japan Nuclear Cycle Development Institute (JNC) is designing the Low level radioactive Waste Treatment Facility (LWTF) in the Tokai Reprocessing Plant (TRP). The low level liquid waste generated the TRP is separated salt (NaNO
, etc) and radionuclide in liquid treatment process of LWTF. The process can get higher volume reduction than previous bituminization. Based on the engineering tests equal to the liquid treatment process of LWTF, the validity of operational condition in LWTF is evaluated. As the results, it is confirmed that all operational condition in the processes which is Iodine immobilization, Pre-filter filtration, Pre-treatment, Coprecipitation and Ultrafiltration are available.
