Abstract
Passive safety systems are an important feature of currently designed and constructed nuclear power plants. They operate independent of external power supply and manual interventions and are solely driven by thermal gradients and gravitational force. This brings up new needs for performance and reliably assessment. This paper provides a review on fundamental approaches to model and analyze the performance of passive heat removal systems exemplified for the passive heat removal chain of the KERENA boiling water reactor concept developed by Framatome. We discuss modeling concepts for one-dimensional system codes such as ATHLET, RELAP and TRACE and furthermore for computational fluid dynamics codes. Part I dealt with numerical and experimental methods for modeling of condensation inside the emergency condenser and on the containment cooling condenser. This second part deals with boiling and two-phase flow instabilities.
Highlights
In part I passive decay heat removal concepts for GEN III(+) nuclear reactors and the modeling of condensation heat transfer in the emergency condenser and in the containment cooling condenser were extensively reviewed and discussed [1]
Full-height scaled loop of Dodewaard, channels made of glass, 4 electrically heated fuel rods, 4 separate bypass channels, 3 m riser, fuel length 1.95 m
The analytic formulations of the complete complex system require a spatial averaging as a simplification due to their large computational effort
Summary
In part I passive decay heat removal concepts for GEN III(+) nuclear reactors and the modeling of condensation heat transfer in the emergency condenser and in the containment cooling condenser were extensively reviewed and discussed [1]. In general in technical systems with thermal hydraulic circuits, either forced or natural, in which phase change by evaporation takes place, two-phase instabilities may occur and can decisively influence the heat removal and the dynamics of these thermal hydraulic processes. Since these systems have safety-relevant functions such as decay heat removal and shutdown, optimal design and knowledge of the mode of operation is fundamental. The integral codes are used to predict the operating performance of the thermal-hydraulic processes in the nuclear power plant in the case of a wide variety of accident scenarios. The modification and improvement of these integral codes in terms of their analytic approaches and computational effort is and remains a subject of current research
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