New models to enhance sodium-cooled fast reactor safety
The EU-funded project JASMIN (Joint advanced severe accidents modelling and integration for Na-cooled fast neutron reactors) was initiated to provide an integral modelling tool for one of the six Gen-IV technologies, sodium-cooled fast reactors (SFRs). The ASTEC-Na platform builds on the European Accident Source Term Evaluation Code (ASTEC) for severe accidents in light-water reactors (LWRs). The currently available computer codes were developed in the 1980s for previous-generation sodium-cooled fast neutron reactor designs. The new ASTEC-Na code for Gen-IV SFRs is an integrated tool with a modern and flexible architecture that eases the integration of new physical models necessary for advanced designs and specific features. Its development is based on existing modules of the LWR ASTEC software and on models derived from the SCANAIR simulation tool. In addition some improved physical models were developed that account for specific SFR phenomena and the results of recent research. This advanced numerical tool is able to assess the consequences of fuel pin failure on materials relocation and primary system loads. In addition, it estimates the potential chemical and radiological source term, which is due both to the radioactive material produced by the relocation of fission products and the formation of sodium oxide/hydroxide particles that may be accidentally released into the environment. The ASTEC-Na computer code and models address four areas for safety related to sodium: thermal-hydraulics, fuel pin thermo-mechanics, source term and neutronics. In particular, the new models describe heat and mass transfer correlation of sodium within the reactor, clad deformation, fission gas release, fuel thermo-mechanical behaviour, and the physical and chemical synthesis of sodium aerosols released to the environment. ASTEC-Na now serves as the European reference code for severe accident scenarios in LWRs and has been extended for use in other water-cooled reactors, including the pressurised water reactor, water-water energetic reactor (VVER), boiling water reactor and CANada Deuterium Uranium (CANDU) reactor. By capitalising acquired knowledge in the ASTEC-Na software and in the experimental database, JASMIN outcomes help to preserve the knowledge produced from more than 40 years of research and development.
Keywords
Sodium-cooled fast reactor, fast neutron reactors, accident scenarios, JASMIN, ASTEC-Na
