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C-stainless steel composites using the improved MPS methodAhmed, Z.*; Wu, S.*; Sharma, A.*; Kumar, R.*; Yamano, Hidemasa; Pellegrini, M.*; Yokoyama, Ryo*; Okamoto, Koji*
International Journal of Heat and Mass Transfer, 250, p.127343_1 - 127343_17, 2025/11
Wakai, Takashi; Ando, Masanori; Okajima, Satoshi; Toyota, Kodai; Onuma, Terumitsu*; Takahashi, Ryoya*; Asayama, Tai
Dai-29-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Yokoshu (Internet), 5 Pages, 2025/06
This paper describes an experimental study for establishing a passive creep-fatigue test technique that mainly utilizes the difference in thermal expansion coefficients of the materials as material surveillance test technique that can be applied to evaluate the structural integrity of the fast reactor components when the components are used beyond the period assumed in the design. Using the test article designed with the aid of a finite element analysis, a long-term creep-fatigue test data has been successfully obtained. In the designing of the test article, it was essential to generate a adequate strain at the gauge portion of the specimen due to the difference of thermal expansion coefficients of the materials, without buckling. After much trial and error, an optimal shape and dimensions of the test article and the cyclic thermal load conditions are established. In the future, miniaturization of the test article for applying the established test technique to the actual nuclear reactors will be required.
C-SS eutectic melting and its relocation behaviourAhmed, Z.*; Sharma, A. K.*; Pellegrini, M.*; Yamano, Hidemasa; Okamoto, Koji*
Arabian Journal for Science and Engineering, 50(5), p.3361 - 3371, 2025/03
Times Cited Count：0 Percentile：0.00(Multidisciplinary Sciences)Mori, Tetsuya; Hazama, Taira; Katagiri, Hiroki*; Ohgama, Kazuya
Nuclear Technology, 211(1), p.143 - 160, 2025/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)The reliability and usefulness of the reaction rate distribution data measured in the prototype fast breeder reactor Monju were examined through a comparison with a calculation using JENDL-4.0, mainly focusing on shielding regions around the reactor core. The 
U(n,f) and 
Ni(n,p) reaction rates sensitive to high-energy neutrons were all judged reliable. The calculation-to-experiment values are slightly worse in the shielding regions, where those for the Ni(n,p) reaction rates were improved by employing JEFF-3.3 instead of JENDL-4.0. A different tendency was observed between the two reactions, probably due to the U(n,f) cross section in the energy range of around 700 eV. The reaction rates of 
U(n,f), 
Pu(n,f), U(n,
), and 
Au(n,) sensitive to the lower energy neutrons were mostly judged reliable. The data in the lower shielding region are less reliable but acceptable for the shielding calculation.
Onoda, Yuichi; Tobita, Yoshiharu; Okano, Yasushi
IAEA-TECDOC-2079, p.215 - 225, 2025/00
The analysis methodologies for the evaluation of unprotected loss of flow accident on sodium-cooled fast reactor in Japan Atomic Energy Agency (JAEA) are briefly explained focusing on the mechanical consequences during expansion phase of the accident. JAEA developed the analysis methodologies for the evaluation of energetics and divided the analysis process into following three: 1) analysis of converting the heat generated into the mechanical energy with SIMMER code, 2) analysis of the structural response of the reactor vessel with AUTODYN code, and 3) analysis of the amount of sodium ejected onto the top shield through the gaps between shield plugs with PLUG code. Pressure-volume relation of the CDA bubble, which is the mixture of gas (fuel, steel vapor and fission gas) and molten core material, obtained by SIMMER calculation is used as the input for structural response analysis with AUTODYN. Pressure history exerted on the lower surface of the top shield obtained by SIMMER calculation is used as the input for PLUG. These analysis codes are validated by simulating the dominant phenomena that significantly affect the results in each calculation. We applied these analysis methodologies developed by JAEA to the reactor case analyses and confirmed their applicability.
Yamano, Hidemasa; Morita, Koji*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 9 Pages, 2024/11
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.
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.
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.
Yamano, Hidemasa; Futagami, Satoshi; Doda, Norihiro; Tagami, Hirotaka; Uchibori, Akihiro; Ogata, Takanari*; Ota, Hirokazu*
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
Matsushita, Kentaro; Ezure, Toshiki; Fujisaki, Tatsuya*; Imai, Yasutomo*; Tanaka, Masaaki
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
An evaluation method of gas entrainment phenomena due to free surface vortices has been developed for the design of a reactor vessel of sodium-cooled fast reactor. The method predicts vortex dimple using the vortex model to the flow field obtained from three dimensional hydraulic analyses of an evaluation area. In this study, the application of adaptive mesh refinement (AMR) method to a water flow experiment in a rectangular channel with advection vortices was examined to create analysis meshes automatically. Transient analyses were conducted using refined meshes obtained by AMR under different initial grid size conditions. Then, the quantities related to vortex formation and the computation cost were compared with the result in a reference mesh with uniformly fine grids. As the result, it was confirmed that the variation of the grid number is possible to use as a criterion to judge the refinement termination in AMR, and the calculated cost of transient analysis can be reduced by AMR.
Yoshikawa, Ryuji; Kikuchi, Norihiro; Tanaka, Masaaki
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
In the study of safety enhancements on advanced sodium-cooled fast reactor, it has been essential to evaluate the influence of buoyancy on pressure drop in a fuel assembly at mixed convection condition during natural circulation under the decay heat removal operation. In this study, the numerical simulations of the 19-rod and 91-rod bundle water experiments at low flow rate conditions were performed for the validation of a thermal-hydraulic analysis code named SPIRAL with the hybrid turbulence model. The influence of buoyancy on the velocity and temperature distributions was analyzed, and the applicability of the hybrid turbulence model to the pressure drop evaluation was investigated by comparison with the experimental friction factors.
Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the application of safety design criteria (SDC) and safety design guidelines (SDG) developed in the Generation-IV International Forum on the natural circulation of sodium to sodium-cooled fast reactors (SFRs) recently designed in Japan.
C-stainless steel using the Moving Particle Semi-Implicit (MPS) MethodAhmed, Z.*; Wu, S.*; Pellegrini, M.*; Okamoto, Koji*; Sharma, A.*; Yamano, Hidemasa
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 14 Pages, 2024/08
The analysis show that once eutectic reaction occurs, the boron diffuses into the stainless steel (SS) wall. Melting initiates at the BC and SS interface, with melt flow following SS cladding penetration. Also, we observed that as temperature rises, a proportional increase in the boron concentration within the melt. The updated MPS method indicated a computational capability of the eutectic reaction model used to effectively analyze control rod eutectic reactions, simulating severe accidents, and its subsequent relocation to understand the effect of BC ingress into the core.
Kikuchi, Shin; Kondo, Toshiki; Doi, Daisuke; Seino, Hiroshi; Ogawa, Kengo*; Nakagawa, Takeshi*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
Yamano, Hidemasa; Takai, Toshihide; Emura, Yuki; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Morita, Koji*; Nakamura, Kinya*; Ahmed, Z.*; Pellegrini, M.*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the project overview and progress of experimental and analytical studies conducted until 2022. A specific result in this paper is to obtain first experimental data of BC-SS eutectic freezing.
Morita, Koji*; Yamano, Hidemasa
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the generalized model developed for these eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as for the reactions that occur between eutectic reaction products in the solid and liquid states and SS or BC. We also describe the thermophysical property model based on thermophysical property data.
Hayashi, Masaaki*; Nakahara, Hirotaka*; Abe, Takashi*; Matsunaga, Suhei*; Miyata, Hajime*; Shirakura, Shota*; Yamano, Hidemasa
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
This paper describes the study of the performance evaluation technology of a heat exchanger between sodium and molten salt and the confirmation of heat transfer improvement measures effects up to FY2023.
Yamano, Hidemasa; Takano, Kazuya; Kurisaka, Kenichi; Kikuchi, Shin; Kondo, Toshiki; Umeda, Ryota; Sato, Rika; Shirakura, Shota*
Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2024/06
This project studies investigation on safety design guideline and risk assessment technology for sodium-cooled fast reactor with the molten-salt heat storage system, development of evaluation method for heat transferring performance between sodium and molten-salt and improvement of the performance, and evaluation of chemical reaction characteristic between sodium and molten-salt and improvement of its safety. This paper describes the effect of sodium-molten salt heat transfer tube failure in addition to the project overview and progress.
Uchibori, Akihiro; Okano, Yasushi
Isotope News, (793), p.32 - 35, 2024/06
The design of a containment vessel in a sodium-cooled fast reactor was optimized from simulation on the hypothetical severe accident including sodium leakage and combustion. The simulation method is one of the base technologies of the design optimization system, ARKADIA. The simulation was performed on the different design conditions including volume of the containment vessel and the safety equipment as optimization parameters. The iterative simulation successfully found that the safety under this accident was kept even in the downsized containment vessel by selecting an effective safety equipment. This study demonstrated that the developed method has basic capability for design optimization in ARKADIA.
