Analysis of the dynamics of a controlled HTR-module-power-plant

The proof of the operational safety of a nuclear power plant constitutes the basis for any safety analysis. For this proof, among other things, it is necessary to make quantitative statements about the operative dynamics of the total system and the loads of single components in their complex interaction among each other. This thesis presents the computer code C.A.S.H., which calculates the dynamic behaviour during operative use and accident scenarios. So the dynamic events in single components of the power plant become accessible to safety analysis. Typical of the new program system is, that the control system and the secondary circuit are added to the description of the primary circuit with the solid temperatures, the thermohydraulics, the neutron dynamics and the steam generator. To simulate the processes of the reactor core and the steam generator C.A.S.H. uses the program codes THERMIX and SIKADE 2. Both are developed at the Instutute for Safety Research and Reactor Technology, KFA Juelich. For simulation of the control system and the secondary circuit new codes have been developed. A main part of this thesis is dedicated to a detailed description of these codes. The program system is used on analysis of dynamic procedures of the Module-Power-Plant. For power transients the suitability of the design of the control system has been proven for the first time. In parallel to the calculations, analytical estimations are made to cover the results of the simulation. The aptitude of the model for description of complex accident transients is proven by calculations of the behaviour of the plant during faulty pull-out of all control rods. Exemplarily it is shown that the design of the control system has an important effect on the accident events during a reactivity accident by faulty pull-out of all control rods