Institute for Neutron Physics and Reactor Technology (INR)

Uncertainty and Sensitivity Analysis Codes

Background

The use of numerical simulation tools for the analysis of safety-relevant phenomena of nuclear power plants in increasing dramatically thanks to the huge and cheap computer power available on desktops and HPC-clusters. For the safety demonstration, safety analysis tools plays a key role and the scope of applications varies from neutronic, thermal hydraulic, thermos-mechanics core analysis, analysis of multidimensional thermal hydraulic phenomena inside the reactor pressure vessel and for plant analysis using best-estimate system thermal hydraulic codes coupled with 3D nodal diffusion solvers.

Goal

The mail goal of the use of uncertainty and sensitivity quantification tools e.g. URANIE, SUSA, DAKOTA, RAVEN, etc. is to quantify the propagation of the uncertainty of  selected input parameters to the output of the applied tools for core analysis and for the analysis of design basis accidents or severe accident. Finally, the limitations and further needs for improvement will be identified.

KIT activities

Many years ago, activities started to get familiar with UQ-tools such as DAKOTA and SUSA and later on URANIE. Latter is being applied to quantify the ASTEC uncertainties in predicting the radiological source term in case of a severe accident sequence for a PWR plant.

 Tool

Description

 SUSA  

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).

DAKOTA

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

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.

Past applications

  • Uncertainty quantification of TRACE for different applications using SUSA and DAKOTA
  • Uncertainty quantification of PARCS for the analysis of BWR REA using SUSA
  • Uncertainty quantification of SubChanFlow for different simulations using SUSA and URANIE

Current applications

  • Uncertainty quantification of the ASTEC severe accident code by simulation QUENCH-tests and a severe accident sequences for a PWR konvoi plant using URANIE and an in-house UQ-tool under development.

Papers

  1. Tomasz Skorek, Agnès de Crécy, Andriy Kovtonyuk, Alessandro Petruzzi, Rafael Mendizábal, Elsad Alfonso, Francesc Reventós, Jordi Freixa,Christine Sarrette, Milos Kyncl, Rostislav Pernica, Jean Baccou, Fabrice Fouet, Pierre Probst, Bubong Chung, TranTranh Tram, Deog Jeon Oh, Alexey Gusev, Alexander Falkov, Yuri Shvestov, Dong Li, Xiaojing Liu, Jinzhao Zhang, Torst Alku, Joona Kurki,Wadim Jäger,Victor Sánchez, Damar Wicaksono, Omar Zerkak, Andreas Pautz; Quantification of th euncertainty of the physical models in the system thermal-hydraulic codes–PREMIUMbenchmark. NED 354 (2019)110199. https://doi.org/10.1016/j.nucengdes.2019.11019
  2. Jaeger, W. & Sánchez Espinoza, V.H., Uncertainty and Sensitivity Study in the Frame of TRACE Validation for Reflood Experiment.  Nuclear Technology Vol.184 (2013) p333-350.
  3. Jaeger, W., Sánchez Espinoza, V.H., Montero Mayorga, F.J. & Queral, C.; Uncertainty and Sensitivity Studies with TRACE-SUSA and TRACE-DAKOTA by means of steady state BFBT Data. Science and Technology of Nuclear Installations. Article ID 610598. 2013.
  4. Jaeger, W., Sánchez Espinoza, V.H., Montero Mayorga, F.J. & Queral, C.; Uncertainty and Sensitivity Studies with TRACE-SUSA and TRACE-DAKOTA by means of transient BFBT Data. Science and Technology of Nuclear Installations. Article ID 565246, 2013.
  5. Jaeger, W., Sánchez Espinoza, V. H.; On the Evaluation of a Fuel Assembly Design by Means of Uncertainty and Sensitivity Measures. Kerntechnik, Vol. 77, No. 5: pp. 324–332, 2012.
  6. Jaeger, W., Sánchez Espinoza, V.H., "Uncertainty and Sensitivity Analysis for the HELIOS Loop within the LACANES Benchmark".  Journal of Energy and Power Engineering Vol.5: pp 515-524, 2011.