Simulation of creep tests with French or German RPV-steel and investigation of a RPV-support against failure

Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its fail-ure time has to be considered for a determination of the loadings on the con-tainment. Numerous experiments have been performed accompanied with ma-terial properties evaluation, theoretical, and numerical work (Rempe, 1993, Theofanous, 1997, Chu 1999).
For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants de-rived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties has been performed. For the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model accord-ing to an approach of Lemaitre is applied.
The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in 3 levels: starting with the simula-tion of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called “tube-failure-experiments” are modeled: the RUPTHER-14 and the “MPA-Meppen”-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. This report deals with the 1D- and 2D-simulations.
An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi5-5 RPV-steels, which are chemi-cally nearly identical. Since these 2 steels show a similar behavior, it should be allowed on a limited extend to transfer experimental and numerical data from one to the other.
After analyzing the FOREVER calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.