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Journal Articles

Investigation on velocity distribution in the subchannels of pin bundle with wrapping wire; Evaluation of Reynolds number dependence in 3-pin bundle

Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 8 Pages, 2020/08

A sodium-cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, the deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, the detailed flow velocity distribution in the subchannel has been obtained by PIV (Particle Image Velocimetry) measurement using a wire-wrapped 3-pin bundle water model. Flow velocity conditions in the pin bundle were set from 0.036 m/s ($$Re$$ = 270) to 1.6m/s ($$Re$$ = 13,500). From the PIV results, the maximum flow velocity was increased by decreasing the $$Re$$ number in the region away from the wrapping wire. Moreover, the PIV measurements by using the 3-pin bundle geometry without the wrapping wire were conducted. From the results, the effect of the wrapping wire on the flow field in the subchannel was understood. There experimental results useful not only for understanding of pin bundle thermal hydraulics but also code validation.

Journal Articles

Measurement of Velocity Field in Five Jets Water Test (FIWAT) for thermal striping in sodium-cooled fast reactor

Aizawa, Kosuke; Kobayashi, Jun; Tanaka, Masaaki; Kurihara, Akikazu; Ishida, Katsuji*; Nagasawa, Kazuyoshi*

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 10 Pages, 2018/11

A conceptual design of an advanced loop type sodium cooled reactor has been carried out in the Japan Atomic Energy Agency (JAEA). Temperature fluctuation is caused by mixing of fluids at different temperature from the control rod channels and the core fuel assemblies, high cycle thermal fatigue may arise on the Core Instrument Plane (CIP) at bottom of the Upper Internal Structure (UIS). In JAEA, 1/3-scaled five jets water tests (FIWAT) have been performed in order to investigate thermal striping phenomena around the CIP. In this study, the velocity field was measured in the mixing area between the jet outlet and the bottom of the structure by using particle image velocimetry (PIV) to compare with the temperature fluctuation characteristics.

Journal Articles

Maximizing $$T_c$$ by tuning nematicity and magnetism in FeSe$$_{1-x}$$S$$_x$$ superconductors

Matsuura, Kohei*; Mizukami, Yuta*; Arai, Yuki*; Sugimura, Yuichi*; Maejima, Naoyuki*; Machida, Akihiko*; Watanuki, Tetsu*; Fukuda, Tatsuo; Yajima, Takeshi*; Hiroi, Zenji*; et al.

Nature Communications (Internet), 8, p.1143_1 - 1143_6, 2017/10

 Times Cited Count:37 Percentile:9.91(Multidisciplinary Sciences)

Journal Articles

Probabilistic risk assessment method development for high temperature gas-cooled reactors, 2; Development of accident sequence analysis methodology

Matsuda, Kosuke*; Muramatsu, Ken*; Muta, Hitoshi*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Takada, Tsuyoshi*; Hida, Takenori*; Tanabe, Masayuki*; et al.

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 7 Pages, 2017/04

This paper proposes a set of procedures for accident sequence analysis in seismic PRAs of HTGRs that can consider the unique accident progression characteristics of HTGRs. Main features of our proposed procedure are as follows: (1) Systematic analysis techniques including Master Logic Diagrams are used to ensure reasonable completeness in identification of initiating events and classification of accident sequences, (2) Information on factors that govern the accident progression and source terms are effectively reflected to the construction of event trees for delineation of accident sequences, and (3) Frequency quantification of seismically-initiated accident sequence frequencies that involve multiplepipe ruptures are made with the use of the Direct Quantification of Fault Trees by Monte Carlo (DQFM) method by a computer code SECOM-DQFM.

Oral presentation

Melting temperature of high-burnup mixed oxide fuels by Re inner capsule method

Hirosawa, Takashi; Sato, Isamu; Miwa, Shuhei; Tanaka, Kosuke; Tanaka, Kenya; Ishida, Takashi*; Sekine, Shinichi*

no journal, , 

no abstracts in English

Oral presentation

Fundamental study of inert matrix fuels adaptable to a fast reactor cycle system, 3; MgO- and Mo-based fuels

Miwa, Shuhei; Osaka, Masahiko; Sato, Isamu; Hirosawa, Takashi; Tanaka, Kosuke; Sekine, Shinichi*; Ishida, Takashi*; Seki, Takayuki*; Kashimura, Naoki*

no journal, , 

no abstracts in English

Oral presentation

Probabilistic risk assessment method development for high temperature gas-cooled reactors, 1; Review of requirements on scenario identification and reliability data development

Muramatsu, Ken*; Muta, Hitoshi*; Matsuda, Kosuke*; Sato, Hiroyuki; Nishida, Akemi; Ohashi, Hirofumi; Itoi, Tatsuya*; Tanabe, Masayuki*

no journal, , 

As part of PRA method development for High temperature Gas-cooled Reactor, domestic and international PRA standards are investigated to identify requirements related to accident scenario assessment and reliability database development. It was concluded that there are three key thing to note for the PRA method development, that is, detail analysis for passive components, integration of mechanistic source term evaluation and supplementation of reliability data.

Oral presentation

Probabilistic risk assessment method development for high temperature gas-cooled reactors, 4; Development of event tree construction and quantification method for accident sequences involving multiple piping ruptures in seismic PRA

Matsuda, Kosuke*; Muramatsu, Ken*; Muta, Hitoshi*; Sato, Hiroyuki; Nishida, Akemi; Itoi, Tatsuya*

no journal, , 

The selection of initiating model for seismic PRA is studied. System analysis are conducted and compared for the case with initiating event hierarchy tree and the case with a multiple branching event tree. The analysis is performed with SECOM2-DQFM code developed by Japan Atomic Energy Agency. As a result of study, we have found the effective classification method for the seismic initiating events with satisfactory accuracy.

Oral presentation

Probabilistic risk assessment method development for High Temperature Gas-cooled Reactors (HTGRs), 9; Consideration on applicability of SECOM 2-DQFM-U code in case of pipe breach accident by seismic

Matsuda, Kosuke*; Muta, Hitoshi*; Muramatsu, Ken*; Otori, Yasuki*; Sato, Hiroyuki; Nishida, Akemi; Itoi, Tatsuya*

no journal, , 

Towards the establishment of evaluation method for accident sequences initiated by seismic, applicability of system reliability analysis code SECOM2-DQFM-U for the sequence is evaluated.

Oral presentation

Study on cooling process of decay heat removal systems in a reactor vessel of sodium-cooled fast reactor by scaled water experiments, 3; Temperature measurement in a reactor vessel simulating operation of reactor vessel auxiliary cooling system

Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu; Nakane, Shigeru*; Ishida, Katsuji*

no journal, , 

Thermal-hydraulic phenomena driven by natural circulation in a reactor vessel was investigated by using scaled model water experiments simulating a reactor vessel in order to enforce of safety and optimize design and operation of decay heat removal systems under normal operation and severe accident conditions. This paper shows temperature measurement results of natural convection flow field in reactor vessel simulating operation of reactor vessel auxiliary cooling system. In addition, it is shown that the affect of dispersal condition of molten fuel on temperature distribution in reactor vessel.

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