Summary
It is of necessity and importance for the simulation of the three‐dimensional thermal hydraulics problem of the pool type fast reactor. However, because of current computing power limitations and the complexity of the reactor core structure, for conventional reactor applications, it is still not possible to directly simulate the entire reactor flow with sufficient fine meshes for detailed pin geometry. Until now, there is a multiscale coupling method which is suitable to deal with this type of simulation challenge. Through the user‐defined function (UDF) of FLUENT, the coupling code FLUENT/KMC‐sub for thermal hydraulic (TH) analysis by coupling the dynamic link library (DLL) complied by the transient subchannel code KMC‐sub is developed by University of Science and Technology of China (USTC). As a code validation case, the steady‐state simulation of a 19‐rod assembly has been carried out by using coupling codes of FLUENT/KMC‐sub, FLUENT and KMC‐sub, and consequently good consistency has been achieved by comparison with experiment results. And coupled code is further tested by comparison with the transient‐state 19‐pin assembly test results of KMC‐sub and FLUENT simulation. This coupling code is then used for TH of M2LFR‐1000 (medium‐size modular lead‐cooled fast reactor) in unprotected loss of flow (ULOF) accident. The transient temperatures of coolant and fuel and multidimensional TH phenomena and safety analysis are presented and discussed in this article.
The simulation of 3D thermal-hydraulic problem for the pool type fast reactors, is one of the necessary and great importance. Most system codes can’t be used to simulate multi-dimensional thermal-hydraulics problems, whereas, the CFD method is suitable to deal with these type of simulation challenges. Based on the CFD method, a neutronics and thermohydraulic coupling code FLUENT/PK for nuclear reactor safety analysis by coupling the commercial CFD code FLUENT with the point kinetics model (PKM) and the pin thermal model (PTM) is developed by University of Science and Technology of China (USTC). The coupled code is verified by comparing with a series of benchmarks on beam interruptions in a lead-bismuth-cooled and MOX-fuelled accelerator-driven system. The variations of transient power, fuel temperature and outlet coolant temperature all agree well with the benchmark results. The validation results show that the code can be used to simulate the transient accidents of critical and sub-critical lead/lead-bismuth cooled reactors. Then this coupling code is used to evaluate the safety performance of MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) at unprotected beam over-power (UBOP) accident, and M2LFR-1000 (Medium-size Modular Lead-cooled Fast Reactor) at the unprotected transient over-power (UTOP) and unprotected loss of flow (ULOF) accident. The transient power, the temperature of coolant and fuel and multi-dimensional flow phenomena in upper plenum and lower plenum are presented and discussed in this paper.
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