interpretation of local flux measurements in subcritical systems and reactivity determination

The determination of the subcriticality level constitutes an important issue in the assessment of the accelerator-driven system technology. For this purpose, the interpretation of flux measurements requires a lumped-parameter model employed in an inverse fashion. This papers addresses the drawbacks of point kinetics in performing such a task. In particular, the problem of the generation of integral parameters is considered, in connection with the use of a shape function and of a projection weight tailored to the neutron flux detector. Furthermore, the question of the generation of the effective source is analysed, and some proposals to modify the time dependence of such a function to account for the time delay at the flux detector are presented and discussed

Dynamics of Fluid Fuel Reactors in the Presence of Periodic Perturbations

The appearance of perturbations characterized by a periodic time behavior in fluid fuel reactors is connected to the possible precipitation of fissile compounds which are moved within the primary circuit by the fuel motion. In this paper the time-dependent response of a critical fluid fuel system to periodic perturbations is analyzed, solving the full neutronic model and comparing the results with approximate methods, such as point kinetics. A fundamental eigenvalue of the problem is defined, characterizing the trend of divergence of the power. Parametric studies on the reactivity insertion, the fuel velocity and the recirculation time are performed, evidencing the sensitivity of the eigenvalue on typical design parameters. Non-linear calculations in the presence of a negative feedback term are then performed, in order to assess the possibility to control a fluid fuel system when periodic reactivity perturbations are involved

Uncertainty quantification in steady state simulations of a molten salt system using polynomial chaos expansion analysis

Uncertainty Quantification (UQ) of numerical simulations is highly relevant in the study and design of complex systems. Among the various approaches available, Polynomial Chaos Expansion (PCE) analysis has recently attracted great interest. It belongs to nonintrusive spectral projection methods and consists of constructing system responses as polynomial functions of the stochastic inputs. The limited number of required model evaluations and the possibility to apply it to codes without any modification make this technique extremely attractive. In this work, we propose the use of PCE to perform UQ of complex, multi-physics models for liquid fueled reactors, addressing key design aspects of neutronics and thermal fluid dynamics. Our PCE approach uses Smolyak sparse grids designed to estimate the PCE coefficients. To test its potential, the PCE method was applied to a 2D problem representative of the Molten Salt Fast Reactor physics. An in-house multi-physics tool constitutes the reference model. The studied responses are the maximum temperature and the effective multiplication factor. Results, validated by comparison with the reference model on 103 Monte-Carlo sampled points, prove the effectiveness of our PCE approach in assessing uncertainties of complex coupled models

Non-intrusive Uncertainty Propagation in the ARC Fusion Reactor through the nemoFOAM Multi-physics Tool

In the framework of the multiphysics analysis of nuclear reactors, it is important to assess the impact of nuclear data uncertainties on relevant thermal-hydraulic quantities like temperature, pressure and mass flow rate. This is particularly important for the safety assessment and for the design verification of fission and fusion systems, through the so-called Best Estimate Plus Uncertainty approach, which qualifies the outputs providing an estimate of their uncertainties. In this work, the uncertainties are propagated from the nuclear data libraries to the thermal-hydraulic quantities of the Breeding Blanket of the Affordable, Robust, Compact fusion reactor thanks to the multiphysics tool nemoFOAM, and employing different uncertainty propagation techniques, like the Total Monte Carlo and the Unscented Transform

Neutronic benchmark of the FRENETIC code for the multiphysics analysis of lead fast reactors

The FRENETIC code is being developed at Politecnico di Torino in the frame of the international effort for the deployment of lead fast reactors technology. FRENETIC is a multiphysics computational tool solving the neutronics and thermal-hydraulics equation at the full-core level, aiming at performing steady-state and time-dependent simulations in different conditions. In the present work, the validation activity of FRENETIC is carried forward by performing a benchmark against a reference computational model for the ALFRED design implemented in Serpent. Different core configurations in FRENETIC and different temperature distributions are considered, performing consistent comparisons between the two codes. All the results obtained show an extremely good agreement between the two models, implying that the ALFRED core can be well characterized by the FRENETIC code. The present study sets the basis for the future application of the code to simulate safety-relevant transients with FRENETIC

Nuclear data uncertainty quantification on PWR spent nuclear fuel as a function of burnup

Nuclear data uncertainty analysis on the spent nuclear fuel inventory was performed on the Takahama-3 NT3G23 assembly, where the sample SF95-4 was irradiated up to a burnup of approximately 36 GWd/ t according to the SFCOMPO benchmark. The cross-section covariance matrices stored in the ENDF/B-VIII.0, JEFF-3.3 and JENDL-4.0u evaluated nuclear data libraries were propagated with the stochastic sampling algorithms implemented in the SANDY code. A comparison of the concentration uncertainty differences obtained using data from the three libraries is reported. Similarities were found with the fuel composition uncertainty results obtained for the Calvert Cliffs MKP109 sample P SFCOMPO benchmark. Such a similarity was also found when comparing concentration uncertainties along the sample irradiation. Therefore, the main contributors to the concentration uncertainty of a number of nuclides were identified at different burnup levels in the two samples. To complement the similarity analysis, a correlation study of the concentration distributions predicted by the two models was performed. The reported results hint a dominance of the common uncertainty propagation mechanisms over the model differences in the determination of concentration uncertainty

A MOC-based neutron kinetics model for noise analysis

A 2-D noise model is implemented in the deterministic reactor code APOLLO3® to simulate a periodic oscillation of a structural component. The Two/Three Dimensional Transport (TDT) solver, using the Method of Characteristics, is adopted for the calculation of the case studies, constituted by a moving detector and control-rod bundle. The periodic movement is built by properly linking the geometries corresponding to the temporal positions. The calculation is entirely performed in the real time domain, without resorting to the traditional frequency approach. A specifically defined dynamic eigenvalue is used to renormalize in average the reactivity over a period. The algorithm is accelerated by the DPN synthetic method. For each cell of the domain, the time values of fission rates are analysed to determine the noise extent. Moreover we propose a systematic approach to the definition of the macroscopic cross sections to be used in dynamical calculations starting from library data. As an aside of our work we have found that even in static calculation this approach can produce significant changes