The field of work of core behaviour at GRS
To analyse the behaviour of the reactor core and its components from a safety-related point of view not only during normal specified operation but also during incidents and under accident conditions is one of the major fields of activity of reactor safety research at GRS.In the specialist field of core behaviour, the scientists work on the following tasks:
• provision of nuclear data for safety-related analyses of core and fuel assembly behaviour
• safety-related analysis of new operating modes of light water reactors (e.g. the assessment of optimisation measures, increase of capacity and burnup, use of MOX fuel and use of new corrosion-proof cladding materials, etc.)
• development of new calculation methods to meet the stricter requirements for accuracy and validity
• review of the safety concepts of new, advanced reactor types, such as Generation-III reactor types, but also of GEN-IV concepts
• systematic registration of uncertainties and sensitivities in nuclear computation systems
The impulses for the work in this field come from the feedback of experience from current processes in German and foreign plants, but new research activities are also triggered by national and international benchmarks and co-operative efforts in this field. The specialist field of core behaviour is divided into the following, closely interrelated topics:
• neutron-physical modelling of stationary and time-dependent phenomena,
• safety-related assessment of core design,
• thermal and mechanical fuel rod behaviour, and
• criticality analyses.
Simulation codes for modelling core behaviour
GRS develops, validates and uses programs and computer codes that allow the analysis and assessment of issues surrounding fuel rod and core behaviour.For example, models are developed for the determination of the criticality behaviour. By coupling thermal hydraulics system codes with 3D core models, calculation codes are provided for the analysis of transients and accidents (e.g. reactivity accidents or loss-of-coolant accidents). These coupled codes allow very exact simulation of the full plant behaviour under normal conditions and in accidents.
The codes are made available to the public. Among others, however, the calculation results are also used in the federal oversight of the nuclear power plants. They form the basis of assessments of the core design of reactors, also with a view to further capacity and burnup increases, as well as of the assessment of new cladding materials.