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powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
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)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)Nishiyama, Nariaki; Nakajima, Toru; Goto, Akira*; Hakoiwa, Hiroaki; Nagata, Mitsuhiro; Shimada, Koji; Niwa, Masakazu
Earth and Space Science (Internet), 11(6), p.e2023EA003360_1 - e2023EA003360_15, 2024/06
Earthquakes with magnitudes of 6-7 have been reported even in various active tectonic settings where fault deformation topography have not been detected. Therefore, delineating concealed active faults generating such earthquakes is necessary to reduce earthquake damage; however, few studies exist to provide its clues regarding such faults. The 1984 Western Nagano Earthquake in Japan was a main shock with a magnitude of Mj 6.8 and a depth of 2 km at the source. Solid bedrocks are well-exposed in the earthquake source region; however, no surface rupture have been identified, and the active fault is known to be concealed. In this study, we collected data on striations observed in fractures by geological survey around the source area of the 1984 Western Nagano Earthquake. Using the collected data, the multiple inverse method was used to estimate the stresses that affected the striation formation. Consequently, stresses similar to acting faults in this area were detected in minor faults around the known concealed active fault. This suggests that the minor faults might be part of the damage zone that has been developed around the concealed active fault. Some minor faults were recognized in Quaternary volcanic rocks, confirming that they experienced displacements recently. This study indicates the possibility of detecting concealed active faults in the bedrock by geological survey.
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.
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.
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.
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.
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.
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)Thwe Thwe, A.; Kadowaki, Satoshi; Nagaishi, Ryuji
Journal of Nuclear Science and Technology, 60(6), p.731 - 742, 2023/06
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)In this study, we performed numerical calculations of unsteady reaction flow considering detailed chemical reactions, investigated the unstable behavior of hydrogen-air dilute premixed flame due to intrinsic instability, and clarified the effects of unburned gas temperature and pressure. I made it. The unstable behavior of the flame in a wide space was simulated, and the burning rate of the cellular flame was obtained. Then, the effects of heat loss and flame scale on flame unstable behavior were investigated. The burning velocity of a planar flame increases as the unburned-gas temperature increases and it decreases as the unburned-gas pressure and heat loss increase. The normalized burning velocity increases when the pressure increases and heat loss becomes large, and it decreases when the temperature increases. This is because the high unburned-gas pressure and heat loss promote the unstable behavior and instability of flame.
Yamada, Taichi; Watanabe, Kaho; Suzuki, Soichiro; Kawabata, Kuniaki
Keisoku To Seigyo, 62(5), p.268 - 271, 2023/05
In Fukushima Daiichi Nuclear Power Station (FDNPS) emergency response and decommissioning, high radiation or unknown environments significantly restrict human workers' activity. Thus, a remotely controlled robot is essential to operate in such an environment instead of human workers. However, remote robot operation is not easy, and it is required to understand the robot's capability, that is, what/how the robot can do on the site. Therefore, robot evaluation method development is important for remote robot operation in disaster sites. We survey the required capabilities for a remotely controlled robot from the remote operation cases in FDNPS and develop test methods to evaluate the capabilities. This paper introduces the survey of FDNPS remote operation cases and the test method development.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
The objective of this study is to develop an effectiveness evaluation methodology of the measures for improving resilience of nuclear structures against excessive earthquake by applying the failure mitigation technology. This study regarded those measures for improving resilience of important structures, systems, and components for safety to enlarge their seismic safety margin. To evaluate effectiveness of those measures, seismic core damage frequency (CDF) is selected as an index. Reduction of CDF as an effectiveness index is quantified by applying seismic PRA technology. Accident sequences leading to loss of decay heat removal are significant contributor to seismic CDF of sodium-cooled fast reactors (SFRs), and those sequences result in core damage via ultra-high temperature condition. This study improved the methodology to evaluate not only the measures against shaking due to excessive earthquake but also the measures at the ultra-high temperature condition. To examine applicability of the improved methodology, a trial calculation was implemented with some assumptions for a loop-type SFR. Within the assumption, the measures for improving resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. Through the applicability examination, the methodology for the effectiveness evaluation was developed successfully.
Kurisaka, Kenichi
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
This study aims to understand a time trend of the occurrence rate of steam generator (SG) tube leak in the existing sodium-cooled fast reactors (SFRs) based on the observed data. The target on SFRs in the present paper is Phenix in France and BN600 in Russia. From the open literature review, we investigated the number of tube-to-tube plate weld, the number of tube-to-tube weld, heat transfer area of tube base metal, operating time of SGs, dates when SG tube leak occurred, leaked location, corrective action after tube leak such as replacement of leaked module. Based on these observed data, time to leak is estimated and then time trend of the occurrence rate of SG tube leak for each of the above-mentioned parts was quantitatively analyzed by the hazard plotting method. As a result, the rate of leak at tube-to-tube weld in Phenix shows increase with time due to probable cause of cyclic thermal stress in a short term. As for a long-term trend, the rate of tube leak in both Phenix and BN600 SGs indicated decrease with time probably thanks to improvement in welding and in SG operating condition and to removal of initial failure.
Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Kamide, Hideki
Handbook of Generation IV Nuclear Reactors, Second Edition, p.173 - 194, 2023/03
Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. JAEA contributes to Chapter 5; Sodium-cooled Fast Reactors (SFRs) and Chapter 12; Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan. Major characteristics and current technology developments including safety enhancement were described in Chapter 5. Chapter 12 shows design activities of SFR. Innovative technology developments, and update of the Japan sodium-cooled fast reactor design with lessons learned from the TEPCO Fukushima Daiichi NPP accident.
Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Nagasawa, Kazuyoshi*; Kurihara, Akikazu; Tanaka, Masaaki
JAEA-Research 2022-009, 125 Pages, 2023/01
JAEA-Research-2022-009.pdf:29.22MB
The design studies of an advanced loop-type sodium-cooled fast reactor (Advanced- SFR) have been carried out by the Japan Atomic Energy Agency (JAEA). At the core outlet, temperature fluctuations occur due to mixing of hot sodium from the fuel assembly with cold sodium from the control rod channels and radial blanket assembly. These temperature fluctuations may cause high cycle thermal fatigue around a bottom of Upper Internal Structure (UIS) located above the core. Therefore, we conducted a water experiment using a 1/3 scale 60 degree sector model that simulated the upper plenum of the advanced loop-type sodium-cooled reactor. And we proposed some countermeasures against large temperature fluctuations that occur at the bottom of the UIS. In this report, we have summarized that the effect of the countermeasure structure to mitigate the temperature fluctuation generated at the bottom of UIS is confirmed, and the Reynolds number dependency of the countermeasure structure and the characteristics of the temperature fluctuation on the control rod surface.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 13 Pages, 2023/00
Times Cited Count：1 Percentile：63.33(Nuclear Science & Technology)Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12
Times Cited Count：1 Percentile：27.23(Nuclear Science & Technology)Inagaki, Satomi; Hayakawa, Misa
Nihon Genshiryoku Gakkai-Shi ATOMO
, 64(12), p.686 - 690, 2022/12
no abstracts in English
