TAILIEUCHUNG - STEAM GENERATOR SYSTEMS: OPERATIONAL RELIABILITY AND EFFICIENCY_2

The book is intended for practical engineers, researchers, students and other people dealing with the reviewed problems. We hope that the presented book will be beneficial to all readers and initiate further inquiry and development with aspiration for better future. The authors from different countries all over the world (Germany, France, Italy, Japan, Slovenia, Indonesia, Belgium, Romania, Lithuania, Russia, Spain, Sweden, Korea and Ukraine) prepared chapters for this book. Such a broad geography indicates a high significance of considered subjects | Part 3 Heat Transfer and Coolant Flow Processes 11 Countercurrent Flow in a PWR Hot Leg under Reflux Condensation Noritoshi Minami1 Michio Murase2 and Akio Tomiyama3 The Kansai Electric Power Company Inc. 2Institute of Nuclear Safety System Inc. 3Graduate School of Engineering Kobe University Japan 1. Introduction Nuclear power plants periodically shut down for plant maintenance and refueling. During a pressurized water reactor PWR plant outage decay heat is removed by residual heat removal RHR systems. The reactor coolant level temporarily needs to be kept around the middle of the primary loop to inspect the steam generator SG tubes and so on. This operation is called mid-loop operation . In some plants a loss of RHR event occurred during the mid-loop operation USNRC 1987 USNRC 1990 . Probabilistic safety assessment PSA studies under plant shutdown conditions have been performed and they confirmed that the loss of RHR cooling during the mid-loop operation is a relatively high risk event for PWR plants. One of the effective methods to cool the reactor core in this event is reflux condensation in which water condensed in a SG flows into the reactor core through a hot leg and cools down the reactor core as shown in Fig. 1. In the reflux condensation steam generated in the core and water condensed in the SG form a countercurrent flow in the hot leg. This phenomenon limits water into the reactor vessel and affects the performance of core cooling. System computer codes as typified by RELAP employ a simplified model to a certain extent to calculate efficiency however the hot leg consists of a horizontal section elbow and inclined section. Hence it is unclear whether the countercurrent flow in the hot leg can be well predicted or not. To improve reliability of the transient analysis we need to understand and model the countercurrent flow in the hot leg. A number of experiments have been made about the countercurrent flow in the hot leg Richter et al. 1978 Ohnuki 1986 .

• Năng lượng
• Generator Tubing
• Temperature Alkaline
• Leaded Solution
• Power Engineering
• Degraded Steam
• Generator Tubes
• Nuclear engineering and technology
• Prognostics for integrity of steam generator tubes
• The general path model
• Nuclear power plants
• Steam Generator Tube Rupture
• Numerical validation of burst pressure estimation equations
• Steam generator tubes with multiple axial surface cracks
• Multiple axial surface cracks
• Giving sufficient confidence