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Journal Articles

JAEA Reports

High-temperature strength of modified type 316 steel for fast reactor fuel before and after neutron irradiation

Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji

JAEA-Technology 2024-009, 140 Pages, 2024/10

JAEA-Technology-2024-009.pdf:8.03MB

For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900$$^{circ}$$C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.

JAEA Reports

Development of the versatile reactor analysis code system, MARBLE3

Yokoyama, Kenji; Hazama, Taira; Taninaka, Hiroshi; Oki, Shigeo

JAEA-Data/Code 2024-007, 41 Pages, 2024/10

JAEA-Data-Code-2024-007.pdf:1.1MB

The third version of the versatile reactor analysis code system, MARBLE3, has been developed. In the development of the former version of MARBLE, object-oriented scripting language Python (Python2) had been used and then the latest version of Python (Python3) was released. However, due to its backward incompatibility, MARBLE no longer worked with Python3. For this reason, MARBLE3 has been fully modified and maintained to work with Python3. In MARBLE3, newly developed analysis codes and newly proposed calculation methods were incorporated, and the user interface was extended and solvers were reimplemented for maintainability, extensibility, and flexibility. In MARBLE3, the three-dimensional hexagonal/triangular transport code MINISTRI Ver.7 (MINISTRI7) and the three-dimensional hexagonal/triangular diffusion code D-MINISTRI are available as the new analysis codes. These codes can be used in the neutronics analysis system SCHEME and the fast reactor burnup analysis system OPRHEUS, which are the subsystems of MARBLE. In addition, the user interface of CBG, a core analysis system embedded in MARBLE, was extended so that the diffusion and transport calculation solvers for the 2-dimensional RZ system of CBG can be used on SCHEME. On the other hand, MARBLE3 has extended the functionality of the burnup calculation solver so that it can use the numerical methods proposed in the papers on the improvement of the Chebyshev rational function approximation method and the minimax polynomial approximation method. From the viewpoint of maintainability, the point reactor kinetics solver POINTKINETICS, which was introduced in MARBLE2, has been newly reworked as the KINETICS solver in MARBLE3.

Journal Articles

Application of the GIF safety design criteria and safety design guidelines on decay heat removal system to next generation sodium-cooled fast reactor in Japan

Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*

Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08

This paper describes the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on decay heat removal system (DHRS) enhancing reliability to sodium-cooled fast reactors (SFRs) recently designed in Japan.

Journal Articles

Application of the GIF safety design criteria and safety design guidelines on reactor shutdown system to next generation sodium-cooled fast reactor in Japan

Yamano, Hidemasa; Futagami, Satoshi; Shibata, Akihiro*

Proceedings of Advanced Reactor Safety (ARS 2024), p.151 - 160, 2024/08

This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the active reactor shutdown system (RSS) to sodium-cooled fast reactors (SFRs) recently designed in Japan.

Journal Articles

Sintering behavior analysis of compacted dry recycled U$$_{0.7}$$Pu$$_{0.3}$$O$$_{2}$$ powder using master sintering curve theory

Nakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi

Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07

Journal Articles

Thinning behavior of solid boron carbide immersed in molten stainless steel for core disruptive accident of sodium-cooled fast reactor

Emura, Yuki; Takai, Toshihide; Kikuchi, Shin; Kamiyama, Kenji; Yamano, Hidemasa; Yokoyama, Hiroki*; Sakamoto, Kan*

Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Journal Articles

The Development of Petri Net-based continuous Markov Chain Monte Carlo methodology applying to dynamic probability risk assessment for multi-state resilience systems with repairable multi-component interdependency under longtermly thereat

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(7), p.935 - 957, 2024/07

 Times Cited Count:1 Percentile:63.33(Nuclear Science & Technology)

Journal Articles

Numerical simulation of accidents involving core damage with integrative severe accident analysis code, SPECTRA

Ishida, Shinya; Uchibori, Akihiro; Okano, Yasushi

Dai-28-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2024/06

no abstracts in English

Journal Articles

Formulation of high-temperature strength equation of 9Cr-ODS tempered martensitic steels using the Larson-Miller parameter and life-fraction rule for rupture life assessment in steady-state, transient, and accident conditions of fast reactor fuel

Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.

Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05

Journal Articles

Numerical study of initiating phase of core disruptive accident in small sodium-cooled fast reactors with negative void reactivity

Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi

Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05

 Times Cited Count:1 Percentile:63.33(Nuclear Science & Technology)

Journal Articles

Validation of the hybrid turbulence model in detailed thermal-hydraulic analysis code SPIRAL for fuel assembly using sodium experiments data of 37-pin bundles

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki

Nuclear Technology, 210(5), p.814 - 835, 2024/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.

Journal Articles

Time-dependent change in occurrence rate of steam generator tube leak in sodium-cooled fast reactors; Phenix and BN-600

Kurisaka, Kenichi

Mechanical Engineering Journal (Internet), 11(2), p.23-00377_1 - 23-00377_14, 2024/04

This study aims to understand the time-dependent change in the occurrence rate of leak from steam generator (SG) tubes in sodium-cooled fast reactors (SFRs). The target SFRs in the present paper are Phenix in France and BN-600 in Russia. By reviewing publicly available literature that show data from the SFRs, we have investigated the numbers of tube-to-tubeplate welds and tube-to-tube welds, heat transfer areas of tube base metal, operating hours of SGs, dates when SG tube leak occurred, locations of leak, and corrective actions taken after tube leak events, such as replacement of the module, in which a leak occurred. Based on these, we have estimated the time to leak and quantitatively analyzed the time-dependent change of the occurrence rates of SG tube leak for each of the above-mentioned parts by hazard plotting method. The results show that the rates of both Phenix and BN-600 decreased over time. For Phenix, this is probably thanks to improved welding and SG operating conditions. For BN-600, it seems that in many cases, the probable cause of the leak was initial defects that developed to failure during the early stage of reactor operation, and that no special countermeasure was taken in the later stages. Therefore, it would be natural to assume that the rate simply decreased over time. The rate of leak at tube-to-tube welds in Phenix shows significant increase in a short term after a certain period of time. This can be caused by thermal stress repeatedly exerted on the materials.

Journal Articles

Fundamental evaluation of hydrogen behavior in sodium for sodium-water reaction detection of sodium-cooled fast reactor

Yamamoto, Tomohiko; Kato, Atsushi; Hayakawa, Masato; Shimoyama, Kazuhito; Ara, Kuniaki; Hatakeyama, Nozomu*; Yamauchi, Kanau*; Eda, Yuhei*; Yui, Masahiro*

Nuclear Engineering and Technology, 56(3), p.893 - 899, 2024/03

Journal Articles

Estimation method for residual sodium amount on unloaded dummy fuel assembly

Kawaguchi, Munemichi; Hirakawa, Yasushi; Sugita, Yusuke; Yamaguchi, Yutaka

Nuclear Technology, 210(1), p.55 - 71, 2024/01

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

This study has developed an estimation method for residual sodium film and sodium lumps on dummy fuel pins in Monju and demonstrated sodium draining behavior through gaps among the pins, experimentally. The amounts of the residual sodium on the surface of the pins were measured using the three-type test specimens: (a) single pin, (b) 7-pin assembly, and (c) 169-pin assembly. The experiments revealed that the withdrawal speed of the pins and improvement of the sodium wetting increased drastically the amounts of the residual sodium. Furthermore, the experiments using the 169-pin assembly measured the practical amounts of the residual sodium in the dummy fuel assembly of short length and demonstrated sodium draining behavior through the dummy fuel assembly. The estimation method includes four models: a viscosity flow model, Landau-Levich-Derjaguin (LLD) model, an empirical equation related to the Bretherton model, and a capillary force model in a tube. The calculation predicted comparable amounts of the residual sodium with the experiments. An uncertain of the sodium wetting effects were close to 1.8 times the estimation values of the LLD model. With this estimation method, the amounts of the residual sodium on the unloaded Monju dummy fuel assembly can be evaluated.

Journal Articles

New market opened up by advanced nuclear reactors (Chapter 3, 4, 5, 7)

Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.

Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10

Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.

Journal Articles

JSME Series in Thermal and Nuclear Power Generation Vol. 3; Sodium-cooled fast reactor development; Joyo, Monju, and demonstration reactor

Ohno, Shuji; Maeda, Seiichiro

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2023/09

Journal Articles

Development of Lagrangian particle method for temperature distribution formed by sodium-water reaction in a tube bundle system

Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08

The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.

Journal Articles

Application study of adaptive mesh refinement method on unsteady wake vortex analysis

Alzahrani, H.*; Sakai, Takaaki*; Matsushita, Kentaro; Ezure, Toshiki; Tanaka, Masaaki

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1262 - 1275, 2023/08

Development of evaluation method for cover gas entrainment by vortices generated at free surface in upper plenum of sodium-cooled fast reactor is required, and an evaluation method by predicting vortices from flow velocity distribution obtained by CFD analysis is developed. In this study, Adaptive Mesh Refinement (AMR) method is examined to improve efficiency of CFD analysis. Initial mesh was refined with two indexes: the first index (Index-1) is when the second invariant, Q, of velocity gradient tensor is negative and the second one (Index-2) is pressure gradient index added to Index-1. As a result of applying AMR method to unsteady vortices system with a flat plate and performing transient analyses with refined meshes, the result of pressure distribution and velocity around the flat plate in mesh using Index-2 was similar to the result of all refined mesh. It was also confirmed that vortices generation and growth was better simulated by refining meshes around separation area.

Journal Articles

The Development of a Multiphysics Coupled Solver for Studying the Effect of Dynamic Heterogeneous Configuration on Particulate Debris Bed Criticality and Cooling Characteristics

Li, C.-Y.; Wang, K.*; Uchibori, Akihiro; Okano, Yasushi; Pellegrini, M.*; Erkan, N.*; Takata, Takashi*; Okamoto, Koji*

Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07

 Times Cited Count:1 Percentile:52.66(Chemistry, Multidisciplinary)

317 (Records 1-20 displayed on this page)