Institute for Neutron Physics and Reactor Technology (INR)

Numeric Tools

LWR Safety: Numeric Simulation Tools

Neutron physical codes:




The NJOY Nuclear Data Processing System

 KANEXT (In-house)

The KArlsruhe Neutronic EXtendable Tool (KANEXT) is a deterministic reactor physics code system developed at KIT/INR since more than 3 decades.  

 KORIGEN  (In-house)

KIT extended version of the ORIGEN2 depletion code


The Standardized Computer Analysis for Licensing Evaluation (SCALE) code system is being developed by ORNL.


Monte-Carlo Neutronentransportcode mit punktweisen Wirkungsquerschnittsdatenbibliotheken und für allgemeine Geometrie.


SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre.



In-house  coupled solutions for

multi-level static core analysis:




High-fidelity coupled code developed at KIT/INR to provide reference solutions for LWR core analysis (static simulations) at Pin- or fuel assembly level taking into account local TH-feedbacks. Due to many innovations, it is able to simulate whole LWR cores at pin level making use of parallel HPC. 



Coupled system for high fidelity simulations of reactor cores taking into account the local TH-feedback. The thermal hydraulic solver (SUBCHANFLOW) is internally coupled to SERPENT2. Validation using experimental data and code-to-code comparison. Very much appropriate to deliver reference solution for fuel assemblies or core problems. Full core analysis at pin-level is also feasible in reasonable time using parallel HPC-environment.         

Reactor dynamic codes:




The Purdue Advanced Reactor Core Simulator (PARCS) is based on multi-group nodal diffusion and SP3 steady state and transient solution for hexagonal and square fuel assembly geometries. It is coupled with RELAP5 and TRACE.


DYN3D is a reactor simulator developed at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) for LWR.


COBAYA3 is a code developed by the Technical University of Madrid (UPM) for the simulation of LWR transients. It is coupled with SUBCHANFLOW, COBRA-TF und FLICA4.

Subchannel Codes:





SUBCHANFLOW is a modern FORTRAN95 code developed at the Institute for Neutron Physics and Reactor Technology (INR) of KIT. As working fluid, not only water but also liquid metals (lead, sodium, Pb-Bi) and gases (e.g. Helium) can be considered in simulations. It is coupled with deterministic (DYN3D, PARCS) and Monte Carlo ( MCNP and SERPENT) codes. SUBCHANFLOW is part of the European Nuclear Reactor Simulation Platform NURESIM



TwoPorFlow is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. Coarse Cartesian grids are used to obtain volume-averaged parameters. The application domain of TwoPorFlow includes the flow in standard porous media and in structured porous media such as micro-channels, spent fuel pools or reactor cores of nuclear power plants.


Coolant-Boiling in Rod Arrays Two Fluids (COBRA-TF) is a Thermal/

Hydraulic (T/H) simulation code for Light Water Reactor (LWR) vessel and core analysis. It uses a two-fluid, three-
field modelling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modelling. In the early 80 it was developed by the PNL for the US NRC and later on it had been used and modi
fied by different institutions. Later on, the Reactor Dynamics and Fuel Management Group (RDFMG) of the PSU started to improve, update, modernize and couple it for different applications. It is part of the CASL platform.


Thermal-hydraulic System Codes:




The Reactor Excursion and Leak Analysis Program is a tool for analysing small-break LOCAs and system transients in PWRs or BWRs.


The TRAC/RELAP Advanced Computational Engine is the reference system code of the US NRC for LWR. It is coupled with PARCS.


ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is a thermal-hydraulic system code developed by the German Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the entire spectrum of leaks and transients in PWRs and BWRs.

Severe Accident Codes:




SCDAP/RELAP5 is a mechanistic severe accident analysis code developed at INEL for the analysis of the early phase of severe accidents in LWR.


The ATHLET-CD (Core Degradation) code is the extended version of ATHLET for the simulation of LWR severe accidents where extensive core damage is involved.


The ASTEC (Accident Source Term Evaluation Code) integral code is being jointly developed by IRSN and GRS since many years for severe accident simulation of LWR.


MELCOR is an integrated coarse mesh code developed at LANL to describe the progression of postulated accidents in light-water reactors, as well as non-reactor systems (e.g., spent fuel pool and dry cask).

Coupled Neutronic - Thermal-hydraulic Codes:




RELAP5/PARCS coupled system is based on the PVM-environment. It is widely used for the analysis of design basis accidents and transient of LWR, especially where strong interactions between the core neutronic s and thermal hydraulics play an important role.


TRACE/PARCS is the reference, best-estimate coupled system of the US NRC for the analysis of transients and DBA of LWR Generation II and III. PARCS is fully integrated in the TRACE code having a unique executable.



In-house coupled code developments:




 Fast running multilevel (nodal/cell) coupled code under development at KIT/INR for the direct prediction of safety parameters. An extension of this system by adding a fuel performance code is under development for improved simulation of RIA or LOCA.



Advanced core simulator for pin and nodal level static and transient simulations of LWR cores. For this purpose, DYN3D-SP3 and SUBCHANFLOW were internally coupled at KIT/INR. The needed nuclear data (pin or nodal) are generated with SCALE6 or SERPENT2 and using the in-house utility tools createXSLIB.

Fuel Performance Codes:




FINIX is a fuel solver under development at VTT capable of predict the fuel temperature in a pin for given power and boundary conditions. It is purely based on public correlations and easy to be use by analysts.


TRANSURANUS is the in-house fuel performance code of JRC/ITU at Karlsruhe. The code predicts the physical status of the fuel at different fuel cycle stages. TRANSURANUS simulates the fuel behaviour using e.g. fuel manufacturing and operational data, reactor and fuel-cladding material properties data as well as operational history for both steady-state and transient simulations (RIA, LOCA, etc.).


Uncertainty and Sensitivity Analysis Codes:




Monte Carlo Program to qualify uncertainties of simulation codes, e.g. neutronic, thermal-hydraulic codes, etc. SUSA is developed by the Gesellschaft für Reaktor- und Anlagensicherheit (GRS).


Design Analysis Kit for Optimization and Terascale Application developed by SNL. It is provided with a graphical user interface (SNAP) which is used for the input generation, manipulation and execution of TRACE simulations.


URANIE is developed for uncertainty quantification and data assimilation and it is based on the Root system of CERN (an object-oriented software multiplatform). It integrates a large amount of features enabled by Root and especially, C++ interpreter, SQL databases access, visualisation tools and statistical analysis.