Analysis on Pipe Rupture of Steam Condensation Line at Hamaoka-1, (II) Hydrogen Combustion and Pipe Deformation

スポンサーリンク

概要

• 論文の詳細を見る

• A pipe rupture occurred in the steam condensing line of the residual heat removal system at the Hamaoka Nuclear Power Station Unit-1 on November 7, 2001. As the accident analysis, the three-dimensional hydrogen combustion behavior was first solved with the one-step irreversible overall reaction model, the Magnussen eddy dissipation model and the SIMPLE method. The ignition point was given at the upstream boundary surface of the non-condensable gas region. The temperature and concentration distributions in the pipe, which were obtained from the analysis of the non-condensable gas accumulation, were given as the initial conditions. The analysis result showed that the detonation pressures in the straight pipes were about 120 MPa and the peak pressures at the elbows were 2.0–2.5 times higher than those in the straight pipes, due to reflection and overlapping of the pressure waves. Then, a three-dimensional dynamic response of pipe deformation was analyzed, with the time transients of the pressure distribution as boundary conditions inside the pipes. The result showed that the strain at the elbow above the surface of the accumulated water exceeded the critical strain and the pipe ruptured there. This was generally consistent with the result of actual pipe deformation observed after the accident.

• 社団法人 日本原子力学会の論文
• 2003-12-25

著者

• NAITOH Masanori

  Nuclear Power Engineering Corporation

• Naitoh Masanori

  Advanced Simulation Systems Department Nuclear Power Engineering Corporation Fujita Kanko

• Ohshima Iwao

  Nuclear And Industrial Safety Agency Ministry Of Economy Trade And Industry

• Kasahara Fumio

  Nuclear Power Engineering Corporation

• KUBOTA Ryoko

  Fuji Research Institute Corporation

関連論文

• 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
• 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
• 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
• Analysis on Pipe Rupture of Steam Condensation Line at Hamaoka-1, (II) Hydrogen Combustion and Pipe Deformation
• Analysis on Pipe Rupture of Steam Condensing Line at Hamaoka-1, (I) Accumulation of Non-condensable Gas in a Pipe
• Severe Accident Analysis Code SAMPSON Improvement for IMPACT Project(Special Issue on International Conference on Power and Energy System)
• ICONE11-36048 Development of Boiling Transition Analysis Code TCAPE-INS/B based on Mechanistic Methods for BWR Fuel Bundles : Models and verification with the boiling transition experimental data
• The Multi-Fluid Multi-Phase Subchannel Analysis Code KAMUI for Subassembly Accident Analysis of an LMFR
• Comparison of sodium evaporations measured under condition of fog formation.

もっと見る 閉じる

スポンサーリンク