ASTEC (Accident Source Term Evaluation Code) allows the numerical simulation of phenomena that could occur in a nuclear reactor in case of severe accidents. It is used to analyse the safety of nuclear reactors and to define actions that prevent or mitigate the consequences of such accidents and improve protection of population in case of accident. This integral code is commonly developed by IRSN and GRS. IRSN is developing a special version for sodium cooled systems to investigate the safety features of fast reactors, which can be used for transmutation of minor actinides. ASTEC–Na is validated within the European programme JASMIN based on separate effects in the CABRI reactor.Both ASTEC and ASTEC-Na are used to simulate the complete scenario of a hypothetical accident in LWR or SFR from the initial event to the final situation of long-term coolability, containment integrity, and fission products retention or release to environment. It consists of a number of modules that can partly be used as stand-alone codes and has been developed in a common effort of IRSN and GRS.
BACCHUS can be used for three-dimensional investigations of single- and two-phase flow conditions thermal-hydraulics in a fuel element with a hexagonal pin arrangement and intact geometry for nominal and accident conditions. The code is based on the porous body concept. Details that may be important for fuel elements in a reactor or experimental setups, including local blockages can be treated locally. For wire wraps, a simplified model is available. Heat conduction in the pins can be calculated one- and three-dimensionally. The code is verified for a large range of applications for sodium and water as a coolant.
EBSILON®Professional is a simulation system for thermodynamic cycle processes that is used for plant planning, design and optimization. EBSILON®Professional supports experts in the planning, from the feasibility study right up to the detailed design of the plant. Owing to the high flexibility of the system and the universality of the approach, any thermodynamic cycle can be modeled: from conventional power plants, nuclear and solar power plants right up to desalination plants, fuel cell applications as well as user-specific processes - there are no limits to the modeling options.
RELAP5 is a one-dimensional thermal-hydraulic "best estimate" system code, developed for USNRC by former INEL, to analyze the reactor coolant systems under normal operation and design basis (i.e. Loss Of Coolant Accidents, LOCA) conditions. It is based on a non-homogeneous and non-equilibrium model for the two-phase system that is solved by a fast, partially implicit numerical scheme to permit economical calculation of system transients. The code includes many generic models allowing to simulate general thermal-hydraulic systems like pumps, valves, heat releasing or absorbing structures, reactor point kinetics, electric heaters, and control system logic elements. At our institute, the code has been significantly improved with respect to reflood situations with high surface temperatures. Its validation includes selected experiment from the QUENCH, the NEPTUN, the FLECHT, and the PKL program. The code development is frozen, TRACE being the follow-on code of the USNRC.
SASSFR is used to analyse complex sequences, severe accidents, and limiting events in Liquid Metal Cooled Reactors (LMRs). The code has originally been developed as SAS4A at ANL (USA) and substantially improved and extended in our institute. The versions SAS-SFR and SAS-LFR are designed to perform deterministic analysis of severe accidents in sodium (SFRs) or lead/lead-bismuth (LFRs) cooled fast reactors with mixed oxide fuels. Detailed mechanistic models describe both the steady state operation and accident conditions caused e.g. by protected and/or unprotected loss of coolant flow accidents or reactivity insertion accidents. SAS-SFR has been extensively qualified with a large variety of results from experiments of the various CABRI programmes, where in-pile tests, in a reactor typical, well-instrumented test environment have been performed, mainly with a single pin.
The in-house codes system SIM-family - SIM-ADS, SIM-LWR, SIM-SFR, SIM-LFR, SIM-GFR, SIM-MSR, is a PC-based; multi node point kinetic model that describes the nuclear and thermal-hydraulic characteristics of both critical and sub-critical reactor cores.
The codes system SIM-family has been used to perform the transient analysis of both the LBE-cooled and He-cooled PDS-XADS ADS designs during the PDS-XADS project, the EFIT-Pb, EFIT-He and XT-ADS designs during the EUROTRANS project, Pb-cooled LFR design during the ELSY project, He-cooled ETDR design during the GCFR project, Na-cooled SFR design during the CP-ESFR project, critical and sub-critical FASTEF reactor designs during the CDT project, He-cooled ALLEGRO and GFR designs during the GoFastR project, Pb-cooled ALFRED and ELFR designs during the LEADER project, Na-cooled ASTRID design during ESNII+ project, LBE-cooled MYRRHA design during the MAXSIMA project, as well as the AMSTER molten salt reactor design during the MOST molten salt project and MSFR design during the EVOL project. The codes system SIM-family has been tested and validated extensively against actual LWR plant data (plant transient data) for both PWRs and BWRs, Superphenix (SPX1) plant data, and by code to code comparisons of results obtained using various releases of RELAP5, SAS4A, SAS-SFR, TRAC, TRACE, SIMMER and other large transient code systems.
The SCDAP/RELAP5 (S/R5) code package, developed for USNRC by former INEL, is composed of the thermal-hydraulic code RELAP5 (see above) and the Severe Core Damage Analysis Package (SCDAP) to describe the material behaviour in the core of a Light Water Reactor (LWR) under severe accident conditions. In our institute, is has been verified against several tests of the CORA and QUENCH out-of-pile series and the in-pile PHEBUS FPT program. Since the code is no longer supported and developed by USNCR, the in-house version is frozen.
TRACE is the follow-on code of RELAP5 of the USNRC, used in this way in our institute. As a special feature, it has a three-dimensional thermal-hydraulic model for the vessel. As part of the international CAMP-Program of the USNRC, it is coupled with the Purdue Advanced Reactor Core Simulator (PARCS) is being qualified by different institutions world-wide.