1.2. INTRODUCTION

The prediction of nuclear reactor thermal-hydraulic behaviour under operational and accidental conditions can potentially be made by

(a) The analysis of the thermal-hydraulic behaviour in smaller scale facilities and extrapolation to real plants, or

(b) The application of computer codes simulating component or system behaviour to plant situations.

The development of nuclear reactor designs has shown a change from small scale experiments in the early days of nuclear reactor development to the application of qualified versions of large computer codes nowadays supported by operational data from real plants. It is generally understood that the assessment under high accidental loads, e. g. in case of loss-of-coolant accidents or even for core melt sequences can only be based on code calculations. Thus the confidence in the code predictions is highly important. It has to be recognized that the requirements on the quality of new experimental data and of code calculations have increased during recent years; this implies the use of actual geometries, materials and thermal-hydraulic boundary conditions — if possible — as well as validation, documentation and uncertainty analysis.