Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.47 - 51, 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.
Bachmann, A. M.*; Nishihara, Kenji; Richards, S.*; Abe, Takumi; Feng, B.*
Progress in Nuclear Science and Technology (Internet), 8, p.11 - 16, 2025/09
Nishihara, Kenji; Sugawara, Takanori; Fukushima, Masahiro; Iwamoto, Hiroki; Katano, Ryota; Abe, Takumi
Progress in Nuclear Science and Technology (Internet), 8, p.185 - 190, 2025/09
A pilot plant for the accelerator-driven system is proposed as a scaled-down version of a lead-bismuth cooled ADS with 800 MW thermal output for transmutation of minor actinides. In this article, the design policy of the pilot plant is elaborated for each design area: safety, subcritical management, sub-critical core, accelerator, target and in-core components. The conceptual design of the pilot ADS is also presented: 200 MW of thermal output, keff below 0.95, introduction of control rod, 1.0 GeV-10mA accelerator and roughly-estimated size of in-core components.
Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Terao, Tsuyoshi*; Koike, Akifumi*
Progress in Nuclear Science and Technology (Internet), 8, p.116 - 120, 2025/09
Japan Atomic Energy Agency, ANSeeN, and Shizuoka University has been conducted a joint-research to develop nuclear instrument for High Temperature Gas-cooled Reactor (HTGR) core power distribution. In the project, we developed the ex-core detector system to avoid heigh temperature environment of the HTGR core. The system achieves the measurement by taking advantage of the HTGR core feature of long flight length neutron due to the graphite moderated core with Computed Tomography (CT) technologies. The theory is demonstrated by calculation in HTTR (High Temperature Engineering Test Reactor) and criticality experiment in KUCA (Kyoto University Criticality Assembly). These technologies are expected to be applied to other reactors.
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.
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.
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.
-hexaoctyl nitrilotriacetamide (HONTA) using mixer-settler extractorsBan, Yasutoshi; Suzuki, Hideya*; Hotoku, Shinobu; Tsubata, Yasuhiro
Progress in Nuclear Science and Technology (Internet), 8, p.243 - 247, 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.
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.
Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Abe, Takumi; Nishihara, Kenji
Progress in Nuclear Science and Technology (Internet), 8, p.52 - 57, 2025/09
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.
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.
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
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
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
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.
Iwasa, Toma; Takano, Masahide
Progress in Nuclear Science and Technology (Internet), 8, p.291 - 295, 2025/09
We have developed the external gelation technology for the fabrication of MAs nitride particles in high-performance heterogeneous fuel. Although the particle fabrication technology using external gelation methods has been developed, there has been almost no study targeting MAs nitrides fuel. Previously study suggested that the size of particle was required to be smaller than 250 
m to avoid the degradation of thermophysical properties. The purpose of this study is to optimize the external gelation conditions for spherical gel particle smaller than 500 m because the particle shrank less than half size by calcination and nitridation. The external gelation tests were performed with the viscosity and pressure of dropping solution as parameters. The results show that the smaller particle with higher sphericity was obtained at the higher pressure of 350-500kPa at each viscosity of 30-50cP with positive correlation.
Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Wozniak, N.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ogata, Takanari*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using a single duct from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement caused by thermal bowing as well as the contact load on the duct pad.
