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Nakahara, Masaumi; Senzaki, Tatsuya; Sano, Yuichi; Kato, Masato
Progress in Nuclear Science and Technology (Internet), 8, p.64 - 69, 2025/09
It has been proposed that minor actinides are recovered and reused as nuclear fuel in a fast reactor fuel cycle system. In this study, minor actinides which were recovered from high-level liquid waste derived from irradiated fast reactor fuel in an extraction chromatography process and U and Pu nitrate solution were mixed, and mixed oxide fuel powders were prepared by microwave heating. The characterization of the mixed oxide fuel powders containing minor actinides was evaluated by X-ray diffraction and thermal analysis.
Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09
Yamamoto, Masahiko; Horigome, Kazushi; Goto, Yuichi; Taguchi, Shigeo; Kuno, Takehiko
Progress in Nuclear Science and Technology (Internet), 8, p.387 - 392, 2025/09
Flush-out of Tokai Reprocessing Plant was completed in February, 2024 for the preparation of process decontamination. Since remaining nuclear materials were contained in main process, purpose of the activities was to recover nuclear materials by transferring them to the highly radioactive liquid waste storage tank and converting uranium solution into uranium trioxide, and to rinse all related processes with nitric acid and water. During this activity, analysis of U and Pu was conducted for nuclear material control and accountancy. Appropriate analytical methods such as isotope dilution mass spectrometry, gravimetric method, spectrophotometry and alpha-ray counting methods were selected depending on the status of flush-out. Also, gamma-ray emitted nuclides in rinsing solution were detected by highly purified germanium detector. This paper describes analysis and its results related to flush-out activities of decommissioning in reprocessing plant.
Sato, Rika; Kondo, Toshiki; Umeda, Ryota; Kikuchi, Shin; Yamano, Hidemasa
Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09
In a sodium-cooled fast reactor (SFR) coupled to thermal energy storage (TES) system, the reaction between nitrate molten salt as thermal energy storage medium and sodium (Na) as reactor coolant might occur under postulated accidental conditions. Thus, the reaction behavior of Na-nitrate molten salt is one of the important phenomena in terms of safety assessment of the SFR with TES system. In this study, reaction experiments on Na-solar salt were performed. It was found that Na-solar salt reaction occurred after the NaNO
-KNO eutectic melting. Based on the measured reaction temperature, the kinetic parameters and rate constant were obtained and compared with the sodium-water reaction. From the results of kinetic analysis, it could be assumed that Na-solar salt reaction occurs in the time frame of the accident such as the failure of heat transfer tube of sodium-molten salt heat exchanger.
Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.47 - 51, 2025/09
Sato, Junya; Takahashi, Yuta; Sunahara, Jun*; Saito, Toshimitsu*; Yoshida, Yukihiko; Sone, Tomoyuki; Osugi, Takeshi
Progress in Nuclear Science and Technology (Internet), 8, p.307 - 312, 2025/09
Kanno, Ikuo; Okumura, Keisuke; Matsumura, Taichi; Riyana, E. S.; Terashima, Kenichi; Sakamoto, Masahiro
Progress in Nuclear Science and Technology (Internet), 8, p.343 - 346, 2025/09
For the estimation of Cs-137 contamination distribution in the gap of shield plugs of the Fukushima Daiichi Nuclear Power Plant Unit 2 with the measurement from its operation floor, a method is proposed using a pinhole camera for gamma-rays. The feasibility is discussed by deterministic calculations.
Nakamura, Satoshi; Suzuki, Hideya*; Ban, Yasutoshi; Ohashi, Akira*
Progress in Nuclear Science and Technology (Internet), 8, p.228 - 232, 2025/09
To reduce volume and radiotoxicity of high-level radioactive waste, JAEA has been developing a separation process recovering minor actinides (MA) from high-level radioactive liquid-waste. In the separation process, separating trivalent MA such as Am and Cm, from RE is challenging due to their similar chemical properties. In this study, extraction properties of three different glycine-based amic-acid-type extractants for MA(III) and RE(III) were studied by a single-stage batch method. The results revealed that the distribution ratios of all metal ions increased with an increase of equilibrium pH, and all extractants showed higher distribution ratios for Am than for RE.
Toigawa, Tomohiro; Tsubata, Yasuhiro; Kumagai, Yuta; Ban, Yasutoshi
Progress in Nuclear Science and Technology (Internet), 8, p.286 - 290, 2025/09
We propose a simple process simulation methodology that uses readily available information about radiation impact. A process simulation was conducted for a minor actinides (MA) separation process while considering the degradation of extraction ability by radiolysis. The simulation provided a processing limit of MA and enabled the evaluation of radiation stability.
Bachmann, A. M.*; Nishihara, Kenji; Richards, S.*; Abe, Takumi; Feng, B.*
Progress in Nuclear Science and Technology (Internet), 8, p.11 - 16, 2025/09
Aihara, Haruka; Hinai, Hiroshi; Shibata, Atsuhiro; Tomita, Sayuri*; Koma, Yoshikazu
Progress in Nuclear Science and Technology (Internet), 8, p.324 - 328, 2025/09
Pu and Am contained in the contaminated water at Fukushima Daiichi Nuclear Power Station is concerned to contaminate inside the buildings concrete. To understand or estimate the state of contamination, investigation on contamination mechanisms have become quite important. Therefore, the distribution ratio of Pu and Am to cement paste and aggregates was obtained by experiments. Cement paste and aggregate were immersed in Pu and Am solution to obtain distribution ratio. Those of Pu and Am to cement paste was high values, suggesting that they have sorbed and accumulated in the building concrete.
Takano, Kazuya; Kurisaka, Kenichi; Yamano, Hidemasa
Progress in Nuclear Science and Technology (Internet), 8, p.82 - 85, 2025/09
As part of the development of risk assessment technologies for sodium-cooled fast reactor coupled to thermal energy storage (TES) system with sodium-molten salt heat exchanger (HX), the database on the number of heat transfer tube failure in molten salt HX and the exposure time of concentrated solar power (CSP) system to molten salt was surveyed on the basis of reported practices of accidents in the existing CSP systems with the molten salt TES system. Using the database, a risk assessment methodology of the heat transfer tube failure frequency was also studied with a Bayesian estimation method to obtain a risk insight for improving the HX.
Arai, Yoichi; Goto, Yasuhiro; Watanabe, So; Agou, Tomohiro*; Arai, Tsuyoshi*; Katsuki, Kenta*; Fukumoto, Hiroki*; Hoshina, Hiroyuki*; Seko, Noriaki*
Progress in Nuclear Science and Technology (Internet), 8, p.329 - 332, 2025/09
Sasaki, Yuji; Kaneko, Masashi; Kumagai, Yuta; Ban, Yasutoshi
Progress in Nuclear Science and Technology (Internet), 8, p.202 - 204, 2025/09
Two extractants and a masking agent of TODGA (TetraOctyl-DiGlycolAmide), ADAAM (AlkylDiAmideAMine), and DTBA (DiethyleneTriamine-triacetic-BisAmide) were developed in JAEA. TODGA can extract both trivalent actinides (An) and lanthanides (Ln), DTBA may separate An from Ln, and ADAAM has high separation factor (SF: 6) for Am/Cm. The suitable conditions for the extraction, separation and isolations of An from Ln are investigated using these reagents. In this work, we show the basic information on extraction behavior of An and Ln using TODGA, DTBA and ADAAM and propose the suitable aqueous and the organic conditions for An+Ln extraction, An/Ln separation and Am/Cm separation.
Hasegawa, Kenta; Ambai, Hiromu; Takahatake, Yoko; Watanabe, So; Nakamura, Masahiro; Sano, Yuichi; Takeuchi, Masayuki
Progress in Nuclear Science and Technology (Internet), 8, p.248 - 251, 2025/09
Aihara, Haruka; Watanabe, So; Kitawaki, Shinichi; Kamiya, Yuichi*
Progress in Nuclear Science and Technology (Internet), 7, p.175 - 181, 2025/03
Myagmarjav, O.; Tanaka, Nobuyuki; Noguchi, Hiroki; Kamiji, Yu; Ono, Masato; Nomura, Mikihiro*; Takegami, Hiroaki
Progress in Nuclear Science and Technology (Internet), 7, p.235 - 242, 2025/03
Sato, Hiroyuki; Yan, X.
Progress in Nuclear Science and Technology (Internet), 7, p.293 - 298, 2025/03
Abe, Takumi; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*
Progress in Nuclear Science and Technology (Internet), 7, p.299 - 304, 2025/03
Currently, much research continues on stable energy sources that do not emit CO
in order to achieve a carbon-neutral and sustainable society. Nuclear energy is one of the such sources, and various new reactors and reprocessing technologies are being developed. In order to implement the nuclear fuel cycle with these technologies, a nuclear fuel cycle simulator is required to quantitatively evaluate various quantities, such as the distribution of nuclear fuel materials and the scale of waste loading. For this purpose, NMB4.0 was developed in collaboration with Tokyo Institute of Technology and Japan Atomic Energy Agency. This code calculates the material balance of 179 nuclides including actinides and fission products (FPs) from the front-end to the back-end and simulates the nuclear fuel cycle in an integrated manner. Unlike other nuclear fuel cycle simulators, the code is capable of performing precise back-end analyses such as the number of radioactive wastes and the scale of the geological repository considering heat generation of waste package under diverse nuclear energy scenario, and is an open source code that runs on Microsoft Excel. By these features, it is possible to quantitatively study nuclear energy utilization strategies with various stakeholders. The presentation will detail the numerical model used in NMB4.0.
Oizumi, Akito; Sagara, Hiroshi*
Progress in Nuclear Science and Technology (Internet), 7, p.331 - 337, 2025/03
Research and development of transuranium (TRU) fuel cycle with accelerator-drive systems (ADSs) transmuting minor actinides separated from the commercial cycles has been continuously conducted to reduce the high-level radioactive waste contained in spent fuel discharged from nuclear power plants. Since the chemical form and composition of the ADS fuels are different from those of the current commercial cycles, it is necessary to examine the inspection goal of the safeguards which are required for the ADS cycle. In this study, the 
of Pu was evaluated assuming the diversion of any items in the ADS cycle in terms of nuclear non-proliferation. They were compared with the 
of the MOX fuel assemblies (fresh and spent fuels) for a conventional boiling water reactor (BWR). The all items in the ADS cycle, regardless of whether they are fresh or spent fuel, were found to have the same of Pu as the spent fuel assembly of BWR-MOX. Additionally, in order to design and evaluate accounting systems for safeguards, the assumed uncertainty (
) values of measurements by operators and verification by regulators for the Pu flow rate in the ADS cycle were derived with reference to the accuracy targets for Pu measurement technology. The derived values were compared with the 1 significant quantity (1SQ), which was generally used as a target value, and the spent fuel standard (equivalent to 5%) set based on the evaluation, respectively. As a result, it was clarified that the both values of Pu measurements by the operator and the regulator exceeded 1SQ but the spent fuel standard was generally achievable.
