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With the FaSTPro (Fast Source Term Prognosis) calculation code for source term prognosis developed by GRS, possible source terms can be predicted at the push of a button in the event of a nuclear power plant accident. The source terms can be transmitted very quickly to authorities or institutions such as the Federal Office for Radiation Protection (BfS) via ready-made forms or data acquisition. For example, it is possible to transfer the data directly to the RODOS decision support system of the BfS, with which the Federal Office, for its part, makes forecasts of the radiological situation.
Until a suitable repository is commissioned, radioactive waste will remain in interim storage. There are 15 storage facilities for high-level radioactive waste in Germany, including 12 decentralised storage facilities directly at the power plant sites and three centralised storage facilities at Ahaus, Gorleben and Lubmin.
Reliable prediction of the characteristics of irradiated light water reactor fuels is needed for many aspects of the reactor operation and for the nuclear fuel cycle.
The GRS simulation program KMACS allows the calculation of core states of pressurised water reactors within one operating cycle, from loading to shutdown. Taking the scheme of the reactor core loaded with fresh and partially spent fuel elements as a basis, the code calculates, for example, detailed temperature and power distributions.
The accident scenarios to be considered in a PSA are characterized by complex interactions over time between the plant behavior, operator actions and stochastic influences.
A central prerequisite for the safe operation of a nuclear installation is that pipes and vessels withstand the occurring loads with a very high degree of reliability. GRS developed the simulation code PROST (PRObabilistic STructural Mechanics) to be able to calculate the occurrence probability and growth of cracks up to possible breaks in metallic components.
QUABOX-CUBBOX provides a detailed analysis of the reactor core behavior based on 3D neutronics models which solve the two-energy group neutron diffusion equations including reactivity feedback effects caused by changes of coolant flow conditions and changes of fuel rod temperatures.
Along with the development and application of computer codes, it has been increasingly recognized that the corresponding computational results are associated with uncertainty due to lack of knowledge on various sources. Therefore, uncertainty and sensitivity analyses are performed to get (1) a quantification of the combined influence of many of these uncertainty sources and (2) a ranking of the individual sources according to their contribution to the uncertainty of the results.
The TESPA-ROD code allows analysing the fuel rod behaviour under various accident conditions, normal operation as well as long-term storage condition. In particular the accident conditions refer to both loss-of-coolant accident (LOCA) and reactivity initiated accident (RIA). The analysis of the fuel rod behaviour under normal operation refers to transients with pellet-cladding (mechanical) interaction (PCI/PCMI) during load follow operation. For long-term storage, the TESPA-ROD code predicts the behaviour for a timescale in the range of 100 years.
As most of the acceptance criteria are local core parameters, safety-relevant quantities, such as fuel rod enthalpy, Departure from Nucleate Boiling Ratio (DNBR), maximum cladding temperature, fuel rod temperature etc., have to be evaluated at local conditions, i.e. at pin cell/subchannel spatial scale, via coupled multi-physics multi-scale neutronics/thermal-hydraulics calculations. This is can be addressed by the transient 3-D fine mesh neutron transport code TORT-TD developed at GRS.