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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.
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.
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.
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.
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,
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.
Hayashi, Masaaki*; Nakahara, Hirotaka*; Shirakura, Shota*; Yamano, Hidemasa
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
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), simple evaluation of heat transfer performance using heat transfer coefficient formula is performed. And Computational Fluid Dynamics (CFD) thermal analyses by STAR-CCM+ with partial HX model are performed to develop evaluation technology. The performance evaluation technology of a HX between sodium and molten salt and the confirmation of heat transfer improvement measures effects is developed.
Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
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.
Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Terao, Tsuyoshi*; Koike, Akifumi*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
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 for 3 years from 2021 supported by "Nuclear Energy System R&D Project" in MEXT. In the project, there are two R&Ds for "Development of ex-core detector" and "Development of in-core detector". The part of "Development of ex-core detector" is reported in this presentation. The "Development of ex-core detector" is innovative technology by virtue of long flight length neutron of graphite moderated HTGR core and Computed Tomography (CT) technologies. These technologies is expected to be applied to other reactors.
Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; Ogata, Takanari*; Wozniak, N.*; Shemon, E.*; 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 of a single duct 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 axial distribution of horizontal duct displacement of a single duct due to thermal bowing and the contact load on the duct pad.
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.
Abe, Takumi; Nishihara, Kenji
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
The robustness of whole of a nuclear fuel cycle (NFC) can be evaluated by a simulation of future operation factors (OFs) of the NFC facilities and mass flow analysis coupled with the simulated OF. In this study, the impact of a reprocessing plant OF on a fast reactor OF was quantified.