Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-Coolant Accident

スポンサーリンク

概要

• 論文の詳細を見る

• The consequence of intentional depressurization of the steam generator (SG) secondary side during a small break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR) was studied using the ROSA-IV Large Scale Test Facility (LSTF). The LSTF is a 1:48 volumetrically scaled full-height model of a Westinghouse-type PWR. The experiment simu- lated a 0.5% cold leg small-break LOCA with total failure of high pressure injection (HPI) system. The SG secondary-side atmospheric relief valve (ARV) was latched open after the core top region was uncovered. Then, the primary pressure closely followed the secondary pressure and dropped to the accumulator (ACC) injection pressure of 4.51MPa before the core became severely overheated. A post test analysis, which was performed using the RELAP5/ MOD2 code with modifications made by the authors, predicted well the primary and secondary side responses during this experiment.

• 社団法人 日本原子力学会の論文
• 1995-02-25

著者

• Kukita Yutaka

  Department Of Nuclear Safety Research Japan Atomic Energy Research Institute

• Kukita Yutaka

  Japan Atomic Energy Research Institute

• Kukita Y

  Nagoya University

• ASAKA Hideaki

  Japan Atomic Energy Research Institute

• Asaka H

  Japan Atomic Energy Research Institute

関連論文

• Dynamic Response of Hot/Cold Liquid Interfaces to Pump Speed Perturbations in a Thermal-Hydraulic Loop Simulating a PIUS-type Reactor
• ROSA/AP600 Testing : Facility Modifications and Initial Test Results
• Secondary-Side Depressurization during PWR Cold-Leg Small Break LOCAs Based on ROSA-V/LSTF Experiments and Analyses
• Core Liquid Level Responses Due to Secondary-Side Depressurization during PWR Small Break LOCA
• Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-Coolant Accident
• Interfacial Friction Factor for High-Pressure Steam/Water Stratified-Wavy Flow in Horizontal Pipe
• Sub- to Supercritical Flow Transition in a Horizontally-Stratified Two-Phase Flow in PWR Hot Legs
• The Simulation Test to Start up the PIUS-Type Reactor from Isothermal Fluid Condition
• Analysis of System Thermal Hydraulic Responses for Passive Safety Injection Experiment at ROSA-IV/Large Scale Test Facility. Using JAERI Modified Version of RELAP5/MOD2 Code.:Using JAERI Modified Version of RELAP5/MOD2 Code
• Simulation of PWR Turbine Trip Transient in ROSA-IV Large Scale Test Facility
• Investigation of break orientation effect during cold leg small-break LOCA at ROSA-IV LSTF.

もっと見る 閉じる

スポンサーリンク