江連 俊樹
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17回
革新炉の統合的な設計評価と最適化に資する ナレッジマネジメントシステムの開発
著者:
近藤 佑樹,橋立 竜太,浜瀬 枝里菜,江連 俊樹,光元 里香,矢田 浩基,江沼 康弘,（JAEA）
発刊日:
公開日:
キーワードタグ:
ARKADIA Fast Reactor Knowledge Base Knowledge Management System Maintenance Optimization Scheme
近藤 佑樹,橋立 竜太,浜瀬 枝里菜,江連 俊樹,光元 里香,矢田 浩基,江沼 康弘,（JAEA）
発刊日:
公開日:
キーワードタグ:
ARKADIA Fast Reactor Knowledge Base Knowledge Management System Maintenance Optimization Scheme
JAEA has started the development of Advanced Reactor Knowledge-and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA) that performs integrated design evaluation and optimization from various viewpoints such as safety, economy, and maintainability using risk information, in order to dramatically improve the development efficiency of innovative reactors. In this paper, we report the concept of a knowledge management system to support integrated design evaluation and optimization,...
英字タイトル:
Development of knowledge management system contributing to integrated design evaluation and optimization of innovative reactors
英字タイトル:
Development of knowledge management system contributing to integrated design evaluation and optimization of innovative reactors