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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.
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.
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.
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
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.
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
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.
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.
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
Ban, Yasutoshi; Suzuki, Hideya*; Hotoku, Shinobu; Tsubata, Yasuhiro
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
A continuous counter-current extraction experiment was performed by mixer-settler extractors to recover minor actinides (MA; Am and Cm) from high-level liquid waste. Using hexaoctyl nitrilotriacetamide (HONTA) as an extractant, 0.17 g of MA was recovered in a MA fraction.
Yamano, Hidemasa; Emura, Yuki; Takai, Toshihide; Kubo, Shigenobu; Quaini, A.*; Fossati, P.*; Delacroix, J.*; Journeau, C.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries. The paper describes major severe accident study results focusing on kinetics of interaction in core material mixtures, physical properties of core material mixtures, high temperature thermodynamic data for the uranium oxide (UO
)-iron (Fe)-boron carbide (B
C) system, experimental studies on BC-stainless steel (SS) kinetics and BC-SS eutectic material relocation (freezing), and BC-SS eutectic and kinetics models for severe accident code systems,
Yamamoto, Masahiko; Horigome, Kazushi; Goto, Yuichi; Taguchi, Shigeo
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
Flush-out activities of Tokai Reprocessing Plant were completed in February, 2024. Since it contained remaining nuclear materials in main process of the facility, purpose of activities was to flush-out them and to rinse with nitric acid solution. This paper describes analysis of nuclear materials related to flush-out activities.
Takeshita, Kenji*; Okamura, Tomohiro*; Nakase, Masahiko*; Nishihara, Kenji; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 2 Pages, 2024/10
Using the dynamic nuclear fuel cycle simulator NMB4.0, the mass balance analysis of the nuclear fuel cycle assuming the introduction of the metal fuel fast reactor in the second half of this century was evaluated. The impact of the introduction of the fast reactor cycle on the back-end including final disposal was discussed.
Bachmann, A. M.*; Richards, S.*; Feng, B.*; Nishihara, Kenji; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
This work demonstrates the value of code verification as an initial step in utilizing fuel cycle simulation. Cyclus and NMB are open-source fuel cycle simulators that provide computational modeling of nuclear fuel cycle alternatives and were chosen by Argonne National Laboratory and Japan Atomic Energy Agency (JAEA), respectively, for a multi-year collaboration on fuel cycle benchmarks. Both are relatively new and can be improved after conducting a rigorous code-to-code comparison. Initial verification of these simulators was performed using a set of hypothetical scenarios for once-through and multi-recycle fuel cycles. The results of this work identify how differences in scenario definitions and the modeling methodologies of the two simulators lead to differences in results in material inventories, mass flows, and other important metrics for fuel cycle assessments.
Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.
Wozniak, N.*; Shemon, E.*; Feng, B.*; Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies.
Nishihara, Kenji; Sugawara, Takanori; Fukushima, Masahiro; Iwamoto, Hiroki; Katano, Ryota; Abe, Takumi
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
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 presentation, the design policy of the pilot plant is presented.
Tagami, Hirotaka; Ishida, Shinya; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu*; Trotignon, L.*; Gubernatis, P.*; Dufour, E.*; Saas, L.*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
JAEA has been developing the SIMMER-V code for severe accident simulations of future sodium-cooled fast reactors, including those with a unique, large-scale heterogeneous core. This paper describes the development framework of SIMMER-V in collaboration with CEA, representative new elements and an example of reactor test calculation.
