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Journal Articles

Investigating eutectic behavior and material relocation in B$$_{4}$$C-stainless steel composites using the improved MPS method

Ahmed, 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

 Times Cited Count:0

Journal Articles

A Study for establishment of passive creep-fatigue test techniques using the difference of thermal expansion coefficients of the materials

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.

Journal Articles

Electronic approach to understand the wettability of surface treated titanium with liquid sodium

Namie, Masanari; Saito, Junichi; Oka, Ryotaro*; Kim, J.-H.*

Vacuum, 234, p.114038_1 - 114038_9, 2025/04

 Times Cited Count:0 Percentile:0.00(Materials Science, Multidisciplinary)

Journal Articles

Radiation heating effects on B$$_{4}$$C-SS eutectic melting and its relocation behaviour

Ahmed, 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:1 Percentile:0.00(Multidisciplinary Sciences)

JAEA Reports

Structural investigation of borosilicate glasses by using XAFS measurement in soft X-ray region, 4 (Joint research)

Nagai, Takayuki; Okamoto, Yoshihiro; Shibata, Daisuke*; Kojima, Kazuo*; Hasegawa, Takehiko*; Sato, Seiichi*; Fukaya, Akane*; Hatakeyama, Kiyoshi*

JAEA-Research 2024-014, 54 Pages, 2025/02

JAEA-Research-2024-014.pdf:7.02MB

XAFS measurements in the soft X-ray region are suitable for evaluating the chemical state of the surface layer of a measurement sample because the X-ray transmittance is low. In this study, the purpose of the study was to confirm the difference between the coagulated surface layer and the inside of the simulated waste glasses by measuring the K-edge of the glass constituent elements boron, oxygen, sodium, and silicon, and the L$$_{3}$$ edge of the waste component cerium. As a result, the B K-edge XANES spectra showed that the proportion of B-O tetracoordinate sp$$^{3}$$ structures (BO$$_{4}$$) on the surface layer of the coagulated glass samples was higher than that on the cut surface inside the glass samples, which is expected to improve the water resistance of the coagulated surface. On the other hand, the O K-edge XANES spectra suggested that the O abundance in the coagulated surface layer was lower than that in the cut surface inside the glass samples, and that alkali metal elements may be concentrated in the coagulated surface layer. However, no difference was observed in the Na K-edge XANES spectra between the coagulated surface layer and the cut surface, and no difference was observed in the Si K-edge XANES spectra between the solidified surface and the inside of glass samples. In addition, the Ce L$$_{3}$$-edge XANES spectra confirmed that the Ce valence in the surface layer of coagulated glass samples were oxidized compared to the inside of glass samples.

Journal Articles

Development of a dissolution method for analyzing the elemental composition of fuel debris using sodium peroxide fusion technique

Nakamura, Satoshi; Ishii, Sho*; Kato, Hitoshi*; Ban, Yasutoshi; Hiruta, Kenta; Yoshida, Takuya; Uehara, Hiroyuki; Obata, Hiroki; Kimura, Yasuhiko; Takano, Masahide

Journal of Nuclear Science and Technology, 62(1), p.56 - 64, 2025/01

 Times Cited Count:1 Percentile:51.66(Nuclear Science & Technology)

A dissolution method for analyzing the elemental composition of fuel debris using the sodium peroxide (Na$$_{2}$$O$$_{2}$$) fusion technique has been developed. Herein, two different types of simulated debris materials (such as solid solution of (Zr,RE)O$$_{2}$$ and molten core-concrete interaction products (MCCI)) were taken. At various temperatures, these debris materials were subsequently fused with Na$$_{2}$$O$$_{2}$$ in crucibles, which are made of different materials, such as Ni, Al$$_{2}$$O$$_{3}$$, Fe, and Zr. Then, the fused samples are dissolved in nitric acid. Furthermore, the effects of the experimental conditions on the elemental composition analysis were evaluated using inductively coupled plasma-atomic emission spectroscopy (ICP-AES), which suggested the use of a Ni crucible at 923 K as an optimum testing condition. The optimum testing condition was then applied to the demonstration tests with Three Mile Island unit-2 (TMI-2) debris in a shielded concrete cell, thereby achieving complete dissolution of the debris. The elemental composition of TMI-2 debris revealed by the proposed dissolution method has good reproducibility and has an insignificant contradiction in the mass balance of the sample. Therefore, this newly developed reproducible dissolution method can be effectively utilized in practical applications by dissolving fuel debris and estimating its elemental composition.

Journal Articles

Phase transitions of sodium peroxide investigated by DSC

Kikuchi, Shin; Koga, Nobuyoshi*

Journal of Thermal Analysis and Calorimetry, 150(1), p.585 - 590, 2025/01

 Times Cited Count:0 Percentile:0.00(Thermodynamics)

Journal Articles

Evaluation of reaction rate distribution for shielding region in the prototype fast reactor Monju

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 $$^{238}$$U(n,f) and $$^{58}$$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 $$^{58}$$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 $$^{238}$$U(n,f) cross section in the energy range of around 700 eV. The reaction rates of $$^{235}$$U(n,f), $$^{239}$$Pu(n,f), $$^{238}$$U(n,$$gamma$$), and $$^{197}$$Au(n,$$gamma$$) 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.

Journal Articles

Analysis methodologies for the evaluation of ATWS accident on SFR in JAEA; Mechanical consequences during expansion phase of the accident

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.

Journal Articles

A Preliminary study for boron mixing effect on severe accident scenario in sodium-cooled fast reactor

Yamano, Hidemasa; Morita, Koji*

Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 9 Pages, 2024/11

Journal Articles

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Multiple duct bowing benchmark

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.

Journal Articles

Validation study on SFR core bowing codes using Joyo ex-core experiment data; Single duct bowing benchmark

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.

Journal Articles

Development of safety design technologies for sodium-cooled fast reactor coupled to thermal energy storage system with sodium-molten salt heat exchanger, 2; Study of sodium-molten salt heat exchanger heat transfer performance

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.

Journal Articles

Study on safety analyses for metal-fueled sodium-cooled fast reactors; Project overview

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

Journal Articles

Validation of thermal-hydraulic analysis code SPIRAL using pressure drop experiments in rod assemblies at mixed convection conditions

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.

Journal Articles

Application of AMR method for numerical analysis of water experiment involving advective vortices

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.

Journal Articles

Applicability of fluorine gas surface treatment to control liquid sodium wettability

Namie, Masanari; Saito, Junichi; Ikeda, Asuka; Oka, Ryotaro*; Kim, J.-H.*

Surfaces (Internet), 7(3), p.550 - 559, 2024/09

Journal Articles

Formation behavior of gaseous iodine from sodium iodide under SFR severe accidental condition

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

Journal Articles

Study on eutectic melting behavior of control rod materials in severe accidents of sodium-cooled fast reactors, 2; Modeling of multi-phase eutectic reaction behavior

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 (B$$_{4}$$C) 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 B$$_{4}$$C. We also describe the thermophysical property model based on thermophysical property data.

Journal Articles

Study on eutectic melting behavior of control rod materials in severe accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2022

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 B$$_{4}$$C-SS eutectic freezing.

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