Kosuke AIZAWA
著者に関連した記事は3件です。
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論文
先進的なナトリウム冷却高速炉の炉心出口部における サーマルストライピング現象に関する水流動試験 (1)制御棒周辺部の温度変動緩和対策の提案
著者:
小林 順,Jun KOBAYASHI,相澤 康介,Kosuke AIZAWA,江連 俊樹,Toshiki EZURE,栗原 成計,Akikazu KURIHARA,田中 正暁,Masaaki TANAKA
発刊日:
公開日:
キーワードタグ:
High Cycle Thermal Fatigue Particle Image Velocimetry Sodium-cooled Fast Reactor Thermal Striping Upper Internal Structure
小林 順,Jun KOBAYASHI,相澤 康介,Kosuke AIZAWA,江連 俊樹,Toshiki EZURE,栗原 成計,Akikazu KURIHARA,田中 正暁,Masaaki TANAKA
発刊日:
公開日:
キーワードタグ:
High Cycle Thermal Fatigue Particle Image Velocimetry Sodium-cooled Fast Reactor Thermal Striping Upper Internal Structure
A design study of an advanced sodium-cooled fast reactor (Advanced-SFR) has been conducted in Japan Atomic Energy Agency (JAEA). Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS). Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A wate...
英字タイトル:
Water Experiments on Thermal Striping Phenomena at the Core Outlet of an Advanced Sodium-cooled Fast Reactor? (1) Proposal of Countermeasures to Mitigate Temperature Fluctuations around Control Rods?
英字タイトル:
Water Experiments on Thermal Striping Phenomena at the Core Outlet of an Advanced Sodium-cooled Fast Reactor? (1) Proposal of Countermeasures to Mitigate Temperature Fluctuations around Control Rods?
論文
先進的なナトリウム冷却高速炉の炉心出口部における サーマルストライピング現象に関する水流動試験 (2)径方向ブランケット燃料集合体周辺部の 温度変動緩和対策の提案
著者:
小林 順,Jun KOBAYASHI,相澤 康介,Kosuke AIZAWA,江連 俊樹,Toshiki EZURE,栗原 成計,Akikazu KURIHARA,田中 正暁,Masaaki TANAKA
発刊日:
公開日:
キーワードタグ:
High Cycle Thermal Fatigue Particle Image Velocimetry Sodium-cooled Fast Reactor Thermal Striping Upper Internal Structure
小林 順,Jun KOBAYASHI,相澤 康介,Kosuke AIZAWA,江連 俊樹,Toshiki EZURE,栗原 成計,Akikazu KURIHARA,田中 正暁,Masaaki TANAKA
発刊日:
公開日:
キーワードタグ:
High Cycle Thermal Fatigue Particle Image Velocimetry Sodium-cooled Fast Reactor Thermal Striping Upper Internal Structure
Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effe...
英字タイトル:
Water Experiments on Thermal Striping Phenomena at the Core Outlet of?an Advanced Sodium-cooled Fast Reactor? (2) Proposal of Countermeasures to Mitigate Temperature Fluctuations?around Radial Blanket Fuel Assemblies?
英字タイトル:
Water Experiments on Thermal Striping Phenomena at the Core Outlet of?an Advanced Sodium-cooled Fast Reactor? (2) Proposal of Countermeasures to Mitigate Temperature Fluctuations?around Radial Blanket Fuel Assemblies?
第14回
研究開発段階発電用原子炉の特徴を考慮した保守管理の提案 (3) 配管支持構造物への適用事例
著者:
相澤 康介,Kosuke AIZAWA,髙屋 茂,Shigeru TAKAYA,近澤 佳隆,Yoshitaka CHIKAZAWA,田川 明広,Akihiro TAGAWA,久保 重信,Shigenobu KUBO
発刊日:
公開日:
キーワードタグ:
maintenance management piping support reactor at R&D stage sodium cooled reactor
相澤 康介,Kosuke AIZAWA,髙屋 茂,Shigeru TAKAYA,近澤 佳隆,Yoshitaka CHIKAZAWA,田川 明広,Akihiro TAGAWA,久保 重信,Shigenobu KUBO
発刊日:
公開日:
キーワードタグ:
maintenance management piping support reactor at R&D stage sodium cooled reactor
One of important mission of nuclear power plants (NPP) at R&D stage is to develop maintenance program for commercial reactors step by step securing safety. Basic principles of maintenance management for NPP at R&D stage were proposed. In this paper, applications for maintenance program on piping support of prototype sodium cooled fast breeder reactor Monju are studied on the basis of the proposed basic principles of maintenance management for NPP at R&D stage. ...
英字タイトル:
Proposal of Maintenance Management of Nuclear Power Plants at R&D Stage by Taking Account of Their Features (3) Application to piping support
英字タイトル:
Proposal of Maintenance Management of Nuclear Power Plants at R&D Stage by Taking Account of Their Features (3) Application to piping support