Institute for Neutron Physics and Reactor Technology (INR)

Methods and Computational Tools

Coupling interface
Shutdowndose
MC and SN coupling
MC and SN coupling
MCMeshTran Coupling
MCMeshTran Coupling
CAD Interface
Coupling Interface
Shutdowndose
MC and SN coupling
MCMeshTran Coupling

Overview

The availability of suitable methods and computational tools is a pre-requisite for any kind of numerical analysis. At INR/NK, the Monte Carlo approach is the standard computational technique for particle transport simulations with the MCNP code as main computational tool. In addition, dedicated methods and computational tools need to be developed to satisfy specific needs for the design and analysis of available and future fusion devices.

Along this guideline, basic development work on methods and computational tools is being conducted in support of project-oriented design analyses. The related activities are embedded in the European Fusion Technology Programme and are supported by EFDA - tasks on ITER, JET, IFMIF and Nuclear Data. 

CAD Interface for MC Particle Transport Codes

McCad is an open source interface programme which automatically translates 3D CAD geometry into a geometry description suitable for Monte Carlo particle transport codes. At present McCad supports the MC codes MCNP5 and Tripoli4. McCad can visualize MCNP models in 3D and make them available for CAD systems

Code system for MC based 3D shutdown dose calculations

During D-T operation of future fusion devices machine components will be activated by neutron radiation generated in the plasma chamber. For safe operation and maintenance it is thus important to be able to predict the induced activation and the resulting shutdown dose rates. 

Coupled MC – Discrete Ordinates computational scheme for 3D shielding calculations

Shielding calculations of advanced nuclear facilities such as the IFMIF neutron source are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. The deep penetration of radiation through bulk shields is a challenge for the Monte Carlo particle transport simulation while approximations are required to model complex geometries by the discrete ordinates method.

MC based sensitivity and uncertainty calculations

Sensitivity and uncertainty analysis is a powerful means to assess uncertainties of nuclear responses in neutron transport calculations and track down these uncertainties to specific nuclides, reaction cross-sections and energy ranges.

Coupling interface code McMeshTran for multiphysics coupled analyses

McMeshTran is an open source interface program for coupled neutronics, thermal hydraulics and structural mechanics analyses. It adopts the volume-weighted interpolation scheme for translating data from MC codes to CFD/FE codes. Currently MCNP5, MCNP6 (unstructured mesh), TRIPOLI, CFX, Fluent and ANSYS Workbench has been interfaced. Also it has been integrated into the SALOME platform to enable user-orientated interactions and visualizations.

 

 

Access to the latest verions of software tools under devepolment at INR is here.