Solving Thermal Hydraulic Safety Issues for Current and New Pressurised Water Reactor Design Concepts: Primary Coolant Loop Test Facility (PKL2) Project – Final Report

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For many years, extensive experimental investigations into the system response of pressurised water reactors (PWRs) under accident conditions have been being conducted at the large-scale Primärkreislauf (PKL) test facility (operated at AREVA NP, Germany) which constitutes a full-height model of the entire reactor cooling system (RCS) and major parts of the secondary side of a PWR.

Since 2001, the PKL Project has been continued in the course of an international project initiated by the Nuclear Energy Agency (NEA) (as recommended in the SESAR report1 ) with the objective of preserving competence in the long term and sufficient infrastructure in the field of reactor safety.

The general objective of the PKL experiments is to contribute to a better understanding of the sometimes highly complex thermal-hydraulic processes involved in various accident scenarios and to allow a better assessment of the countermeasures implemented for accident control and the demonstration of safety margins available in the plants. In addition, the experimental results aim at the application in the validation and further development of thermal-hydraulic computer codes, also called system codes.