UNCERTAINTY ANALYSIS OF SUB-EXERCISES IN UAM-SFR BENCHMARK WITH THE MCS CODE

Yunki Jo,Hyunsuk Lee,Deokjung Lee,Vutheam Dos,Nhan Nguyen Trong Mai,M Margulis,P Blaise

doi:10.1051/epjconf/202124715017

Abstract

Uncertainty analysis in Modelling (UAM) for Design, Operation and Safety Analysis of Sodium-cooled Fast Reactors (SFRs) has been formed by OECD/NEA to assess the effect of nuclear data uncertainties on parameters of interest in SFR analysis. In this paper, sub-exercises of a medium 1000 MWth metallic core (MET-1000) and a large 3600 MWth oxide core (MOX-3600) are tested by a Monte Carlo code MCS to perform uncertainty analysis. Classical perturbation theory and generalized perturbation theory are used to calculate sensitivity coefficients. Uncertainty is calculated by multiplying the sensitivity coefficients and relative covariance matrix from ENDF/B-VII.1 library.

Highlights

Uncertainty analysis in modelling (UAM) for design, operation and safety analysis of Sodium-cooled fast reactors (SFRs) has been formed by OECD/NEA to investigate the effect of nuclear data uncertainties on SFR full core calculations [1]
Uncertainty analysis has been performed to quantify the effect of nuclear data uncertainty on k∞, microscopic and macroscopic cross sections in UAM-SFR benchmark
A Monte Carlo code MCS has been used as an uncertainty quantification tool

Summary

INTRODUCTION

Uncertainty analysis in modelling (UAM) for design, operation and safety analysis of Sodium-cooled fast reactors (SFRs) has been formed by OECD/NEA to investigate the effect of nuclear data uncertainties on SFR full core calculations [1]. The effect of nuclear data uncertainty has been assessed for the sub-exercises in the UAMSFR benchmark. A Monte Carlo code MCS is used as a computational tool for neuron transport calculation and sensitivity/uncertainty (S/U) analysis. MCS [4] is a high-fidelity Monte Carlo particle transport code that has been developed at Ulsan National Institute of Science and Technology (UNIST). It has capabilities of thermal-hydraulics feedback, on-the-fly cross section treatment and depletion calculation. To investigate top contributors to uncertainties, sensitivity coefficients and uncertainty of nuclide cross section will be compared between MET-1000 and MOX-3600 models

UNCERTAINTY QUANTIFICATION METHOD

Pin-cell Models

Fuel Assembly models

Super-cell models

CONCLUSIONS