Search Results 1-17 of 17

  • Prospects based on T-H Roadmap through Communication  [in Japanese]

    Nakamura Hideo

    Journal of the Atomic Energy Society of Japan 61(4), 270-272, 2019

    J-STAGE 

  • Japanese T-H:A Small Step Towards Giant Leap  [in Japanese]

    Miwa Shuichiro

    Journal of the Atomic Energy Society of Japan 61(4), 273-275, 2019

    J-STAGE 

  • Thermal-Hydraulics Technological Strategy Roadmap 2017, An Approach for Continuous Safety Improvement of LWRs  [in Japanese]

    AESJ Thermal-Hydraulics Division with a cooperation from Computational Science and Engineering Division

    <p>　熱流動部会は福島第一原子力発電所(1F)事故の教訓を基にした熱水力分野のロードマップの改訂活動(ローリング)を進め，2017年3月に｢熱水力安全評価基盤技術高度化戦略マップ2017(熱水力ロードマップ2017)｣を策定した。世界最高水準の安全性の実現とその継続的改善を図るため，安全裕度向上策および人材育成に必要なニーズとシーズのマッチングを考慮して選定・詳述された2015年版の …

    Journal of the Atomic Energy Society of Japan 60(4), 221-225, 2018

    J-STAGE 

  • Thermal-Hydraulics technological strategy roadmap that improves safety of LWRs  [in Japanese]

    Thermal-Hydraulics Division of AESJ

    <p>　原子力学会 熱流動部会は，福島第一原子力発電所(1F)事故の教訓を基に熱水力安全評価基盤技術高度化戦略マップ2015(改訂版)を他分野に先駆けてH27年3月に策定した。世界最高水準の安全性を実現しその継続的向上を図るため，深層防護を柱にシビアアクシデントや外的事象の対策を整理し，安全裕度向上策および人材育成に必要なニーズとシーズのマッチングを考慮した上で技術課題を選定し，1F廃 …

    Journal of the Atomic Energy Society of Japan 58(3), 161-166, 2016

    J-STAGE 

  • ICONE23-1046 MODELLING OF LOW-PRESSURE SUBCOOLED FLOW BOILING USING ATHLET CODE

    Li Fei , Meng Zhaocan , Liu Xiaojing , Li Linsen , Shen Feng , Cheng Xu

    … The best-estimate thermal-hydraulic computer code ATHELT (Analysis of Thermal-Hydraulics of Leaks and Transients) is developed by the GRS (Gesellschaft fur Anlagen-und Reaktorsicherheit). …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23(0), _ICONE23-1-_ICONE23-1, 2015

    J-STAGE 

  • An Experimental Investigation of Boiling of Water in a Vertical Tube

    MATSUMURA Kunihito , KAMINAGA Fumito , SUZUKI Masahiro , FUJII Kan-ichi

      In order to realize a quick mass production of PuO<sub>2</sub>/UO<sub>2</sub> (MOX) particles <i>via</i> vaporization of Pu/U mixed nitrate solution, an …

    Transactions of the Atomic Energy Society of Japan 13(2), 51-61, 2014

    J-STAGE 

  • ICONE19-43257 TUBE SUPPORT PLATE BLOCKAGE IN STEAM GENERATORS : INVERSE METHOD FOR THE ESTIMATION OF UNOBSERVABLE PARAMETERS IN A DEPOSIT MODEL

    PRUSEK Thomas , MOLEIRO Edgar , OUKACINE Fadila , TOUAZI Ouardia , ADOBES Andre

    … In order to improve our understanding of those two phenomena, a model for the growth of solid deposits on the secondary side of steam generators has been developed by the R&D Division of EDF. … This model has been implemented in the frame of THYC, which is the EDF reference code for the modelling of two-phase thermal-hydraulic phenomena at the steam generator scale. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19(0), _ICONE1943-_ICONE1943, 2011

    J-STAGE 

  • Present Status and Prospect of the Thermal-Hydraulics Division  [in Japanese]

    Thermal-Hydraulics Division , 山口 彰 , 成合 英樹

    <p>　熱流動部会は，委員会活動や国際会議開催など様々な活動を通じて，原子力分野における熱流動技術の発展，規格・基準の策定，人材育成等に貢献してきた。本稿では，そのような熱流動部会の設立の経緯，主な活動の経緯，関連する技術分野の状況とそれに対する熱流動部会の貢献などについて紹介し，今後の部会活動に対する展望を述べる。</p>

    Journal of the Atomic Energy Society of Japan 51(4), 270-274, 2009-04-01

    J-STAGE  References (20)

  • Computational Fluid Dynamic Studies of the MEGAPIE Spallation Source Target and Safety Vessel

    SMITH Brian L. , DURY Trevor V. , NI Liping , ZUCCHINI Alberto

    Journal of Nuclear Science and Technology 45(12), 1334-1346, 2008-12-01

    References (25)

  • Study on Temperature Fluctuation Characteristics for High Cycle Thermal Fatigue in a Mixing Tee  [in Japanese]

    IGARASHI Minoru , KAMIDE Hideki , TANAKA Masaaki , KIMURA Nobuyuki

    … A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. …

    Transactions of the Japan Society of Mechanical Engineers. Series B. 70(700), 3150-3157, 2004-12-25

    References (18) Cited by (2)

  • Study on Temperature Fluctuation Characteristics for High Cycle Thermal Fatigue in a Mixing Tee  [in Japanese]

    IGARASHI Minoru , KAMIDE Hideki , TANAKA Masaaki , KIMURA Nobuyuki

    … A water experiment for thermal hydraulics in a mixing tee was performed to investigate thermal striping phenomena. …

    TRANSACTIONS OF THE JAPAN SOCIETY OF MECHANICAL ENGINEERS Series B 70(700), 3150-3157, 2004

    J-STAGE 

  • Conceptual design of the horizontal modular storage system for spent nuclear fuel  [in Japanese]

    Minami Ryogo , Nakata Tetsuo , Takahashi Nobuyuki

    横型サイロ貯蔵施設は, 米国で実用化された軽水炉の使用済み燃料乾式貯蔵システムであり, 我が国の使用済み燃料中間貯蔵施設の候補概念の一つであることから, 経済性向上, 日本の安全審査指針・技術基準, 貯蔵燃料の仕様, 耐震条件等との適合性を考慮して, 我が国向けの横型サイロ貯蔵施設の開発を進めてきた． <br>　横型サイロ貯蔵施設の設計に当たっては, 日本の条件に適合させるため, 除熱 …

    Journal of Nuclear Fuel Cycle and Environment 9(2), 115-120, 2003

    J-STAGE 

  • ICONE11-36446 ASSESSMENT OF THE RELAP5-3D CODE Against An IIST 0.5% COLD LEG SBLOCA EXPERIMENT WITH PASSIVE SAFETY INJECTION

    Chang Chin Jang , Young Hua Jiun , Lee Chien Hsiung , Wang Lance L. C.

    … The code predictions include the analysis for primary system pressure, loop flow rate, loop and CMT temperatures, coolant inventory distribution in pressurizer, Accumulator (ACC), and Core Makeup Tank (CMT), mass flow rate in Passive Residual Heat Removal (PRHR) system, and other core thermal-hydraulics. …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003(0), 306, 2003

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  • ICONE11-36057 LAGRANGIAN EXTENSIBLE SIMULATION MODULES FOR FUEL-COOLANT INTERACTION

    Nilsuwankosit Sunchai , Song Jin Ho

    The numerical modules for simulating the motion of the Lagrangian particles and their interactions with the surrounding Eulerian fluids are currently being developed at Korea Atomic Energy Research In …

    The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2003(0), 224, 2003

    J-STAGE 

  • Validation of the MSG (Multi-dimensional Thermal-hydraulics Analysis Analysis Code for Steam Generators) for CFD Modelling of Sodium Heated Steam Generators

    METZ Peter , YOSHIKAWA Shinji

    Journal of Nuclear Science and Technology 38(12), 1126-1132, 2001-12-25

    References (9)

  • Experimental study on PORV break LOCA in PWR plants.

    KAWANISHI Kouhei , NAKAMORI Nobuo , TSUGE Ayao , KODAMA Kenji , KOHRIYAMA Tamio , NAGUMO Hiroichi

    Small break loss-of-coolant accident (LOCA) tests simulating a pressurizer power operated relief valve (PORV) break LOCA were performed using the EOS (Emergency of System) test facility. The break siz …

    Journal of Nuclear Science and Technology 27(2), 133-148, 1990

    J-STAGE  Cited by (1)

  • THE ROLE OF DROPS IN ANNULAR GAS-LIQUID FLOW: DROP SIZES AND VELOCITIES

    Azzopardi B. J.

    Methods of measurement and data obtained for drop sizes and velocity in annular gas/liquid flow have been reviewed. In particular, possible sources of bias of bias in the results have been examined. T …

    JAPANESE JOURNAL OF MULTIPHASE FLOW 2(1), 15-35, 1988

    J-STAGE 

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