Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits

The impact of nuclear data uncertainties is studied for the reactor power and the reactivity during control rod withdrawal transients with reactivity insertions of 0.5$and 0.97$, respectively, for a PWR mini-core model. Multi-group cross sections, the multiplicities of both prompt and delayed neutrons, and fission spectra are varied by the application of the random sampling-based method XSUSA with covariance data of SCALE 6.1 supplemented by JENDL-4.0. The varied multi-group data are used by TRITON/NEWT to generate varied 2-group cross sections, which are then applied in neutron-kinetic/thermal-hydraulic calculations with DYN3D-ATHLET. A significant impact on both the reactivity uncertainty and the power uncertainty is observed. Since the distributional properties of the output time series vary across the problem time, the distribution-free Wilks tolerance limit is applied as a robust uncertainty measure to complex time series patterns. The most contributing nuclide reactions to the power uncertainty are identified via sensitivity analysis. (C) 2019 Elsevier Ltd. All rights reserved

NUCLEAR DATA UNCERTAINTY AND SENSITIVITY ANALYSIS WITH XSUSA FOR FUEL ASSEMBLY DEPLETION CALCULATIONS

Uncertainty and sensitivity analyses with respect to nuclear data are performed with depletion calculations for BWR and PWR fuel assemblies specified in the framework of the UAM-LWR Benchmark Phase II. For this, the GRS sampling based tool XSUSA is employed together with the TRITON depletion sequences from the SCALE 6.1 code system. Uncertainties for multiplication factors and nuclide inventories are determined, as well as the main contributors to these result uncertainties by calculating importance indicators. The corresponding neutron transport calculations are performed with the deterministic discrete-ordinates code NEWT. In addition, the Monte Carlo code KENO in multi-group mode is used to demonstrate a method with which the number of neutron histories per calculation run can be substantially reduced as compared to that in a calculation for the nominal case without uncertainties, while uncertainties and sensitivities are obtained with almost the same accuracy

OECD/NEA intercomparison of deterministic and monte carlo cross-section sensitivity codes using sneak-7 benchmarks

International audienceA sensitivity benchmark exercise was organized within the scope of the Uncertainty Analysisin Modeling (UAM) project of the OECD/Nuclear Energy Agency (NEA) to develop andcompare methods for the sensitivity and uncertainty computations of the effectivemultiplication factor (keff) and the effective delayed neutron fraction (eff). Several solutionswere received using different codes, both deterministic (SUSD3D, SNATCH) and MonteCarlo (TSUNAMI-3D, XSUSA, SERPENT2, MCNP6). In this paper the performances ofseveral codes and methods for the keff sensitivity and uncertainty computations are intercompared. The sensitivity and uncertainty codes were applied to the SNEAK-7A and -7B fastneutron benchmark experiments from the IRPhE database. Good general agreementbetween the sensitivities, both for integral values and sensitivity profiles, was observed

OECD/NEA intercomparison of deterministic and monte carlo cross-section sensitivity codes using sneak-7 benchmarks

International audienceA sensitivity benchmark exercise was organized within the scope of the Uncertainty Analysisin Modeling (UAM) project of the OECD/Nuclear Energy Agency (NEA) to develop andcompare methods for the sensitivity and uncertainty computations of the effectivemultiplication factor (keff) and the effective delayed neutron fraction (eff). Several solutionswere received using different codes, both deterministic (SUSD3D, SNATCH) and MonteCarlo (TSUNAMI-3D, XSUSA, SERPENT2, MCNP6). In this paper the performances ofseveral codes and methods for the keff sensitivity and uncertainty computations are intercompared. The sensitivity and uncertainty codes were applied to the SNEAK-7A and -7B fastneutron benchmark experiments from the IRPhE database. Good general agreementbetween the sensitivities, both for integral values and sensitivity profiles, was observed