Implementation of thermodynamic properties of lead in the TRACE system code towards its integration in a full-scope platform for safety assessments

Lead-cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology roadmap of the Generation IV International Forum (GIF) as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analyses codes.JRC.F.5-Nuclear Reactor Safety Assessmen

Development of a three-dimensional thermohydraulic-neutronic coupling scheme for transient analysis of liquid metal cooled fast reactor technologies using the system code TRACE-PARCS

According to the R&D agenda of the SNETP (Sustainable Nuclear Energy Technology Platform) and Generation IV International Forum (GIF) there is a need to develop computational tools to ensure the safety of new reactor concepts. These computational tools should provide a reliable and accurate way to simulate conditions that may threat the reactor safety. Accident scenarios in a Nuclear Power Plant (NPP) involve 3D phenomena that these tools should be able to simulate. With this aim, a 3D thermo-hydraulic model of a Sodium Fast Reactor (SFR) has been developed for the system code TRACE. In order to take account the neutronic behaviour of the core during such transients, this model has been coupled with a 3D neutron kinetics code (PARCS). These codes, aimed at the simulation of transients of light water reactors, have been adapted to take into account for the specific phenomena involved in fast reactors. The set of cross sections needed for such analysis has been generated using the Monte Carlo code SERPENT. The aim of this paper is to describe the details of such computational structure and to show its results in the analysis of a SFR-type plant. This multi-physics computational structure could be easily tailored for other liquid metal cooled fast reactor technologies.JRC.F.5-Nuclear Reactor Safety Assessmen

Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs Sodium Fast Reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure the reactor safety level. In this line, several organizations developed models able to simulate the complex and specific phenomena involving multi-physics studies that this fast reactor technology requires. In the introduction of this paper the framework of this study is described, the second section describes the plant design and the modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as checking the implementation of all relevant physical phenomena in the respective codes.JRC.F.5-Nuclear Reactor Safety Assessmen

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimized core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.JRC.F.5-Nuclear Reactor Safety Assessmen

Generation IV Reactor Safety and Materials Research by the Institute for Energy and Transport at the European Commission’s Joint Research Centre

To support the drafting, development, implementation and monitoring of European energy and transport policy, the Institute for Energy and Transport of the European Commissions’ Joint Research Centre conducts pre-competitive research in the areas of experimental qualification of advanced fuels and materials as well as simulation and modelling of reactor safety and material performance. The work covers assessments, design optimisation and improvements to the safety and performance of new, innovative reactor systems, materials and instrumentation, in order to meet the EU’s long-term energy needs while respecting sustainability, enhanced safety and economic aspects. The research is linked, and contributes, to related EURATOM Framework Programme projects, Generation IV International Forum (GIF), International Atomic Energy Agency as well as OECD’s Nuclear Energy Agency (OECD NEA) activities. The current paper gives an overview and examples of past, current, and upcoming activities in the areas of reactor safety assessments, advanced fuel irradiation and materials research.JRC.F.5-Nuclear Reactor Safety Assessmen