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Nuclear Power Engineering Corporation | 論文
- 原子炉熱流動の微視的シミュレーション
- P25-05 シビアアクシデント解析コード SAMPSON を用いた実証解析
- P25-04 改良型加圧水型軽水炉炉内流動解析コード改良試験流動試験結果と IMPACT コードによる解析結果との比較
- C103 IMPACT プロジェクトにおけるシビアアクシデント解析コードの開発
- Post DNB Heat Transfer Experiments for PWR Fuel Assemblies
- Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (II) : Evaluation of Corrosive Conditions in PWR Secondary Cooling System
- Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (I) : Major Targets and Development Strategies of the Evaluation Methods
- Deterministic Trigger Model for the VESUVIUS Steam Explosion Code
- Analysis of International Standard Problem No. 45, QUENCH06 Test as FZK by Detailed Severe Accidents Analysis Code, IMPACT/SAMPSON
- ICONE11-36423 Analysis of Fission Product Behaviors in the Phebus-FPT1 Test with IMPACT/SAMPSON Code
- ICONE11-36297 ANALYSIS OF THE THERMAL HYDRAULICS AND CORE DEGRADATION BEHAVIOR IN THE PHEBUS-FPT1 TEST TRAIN WITH IMPACT/SAMPSON CODE
- Analysis of Core Degradation and Fission Products Release in Phebus FPT1 Test at IRSN by Detailed Severe Accidents Analysis Code, IMPACT/SAMPSON
- Effects of Hydrogen Peroxide on Corrosion of Stainless Steel (VI) : Effects of Hydrogen Peroxide and Oxygen on Anodic Polarization Properties of Stainless Steel in High Temperature Pure Water
- Development of the VESUVIUS Code for Steam Explosion Analysis : Part 2 : Verification of Jet Breakup Modeling
- Development of the VESUVIUS Code for Steam Explosion Analysis : Part 1 : Molten Jet Breakup Modeling
- Reflux Condensation Heat Transfer of Steam-Air Mixture under Turbulent Flow Conditions in a Vertical Tube
- Evaluation of Reflux Condensation Heat Transfer of Steam-Air Mixtures under Gas-Liquid Countercurrent Flow in a Vertical Tube
- Proving Test on Seismic Reliability of Piping System with Energy Absorbing Supports : Verification of Seismic Response Analysis Method of Piping System
- Prediction of Dryout Heat Flux Using Film Flow Model
- Critical Power Analysis with Mechanistic Models for Nuclear Fuel Bundles,(I) Models and Verifications for Boiling Water Reactor Application