237 research outputs found
Generalized Perturbation Techniques for Uncertainty Quantification in Lead-Cooled Fast Reactors (dataset and software)
No full textDataset and codes to allow the reproducibility of the results presented in the publication:
N. Abrate, S. Dulla, P.Ravetto, "Generalized Perturbation Techniques for Uncertainty Quantification in Lead-Cooled Fast Reactors", Annals of Nuclear Energy (2021
Erratum: “Radiative heat load distribution on the EU-DEMO first wall due to mitigated disruptions” (Nuclear Materials and Energy (2020) 25, (S2352179120300971), (10.1016/j.nme.2020.100824))
No full textThe publisher regrets for the incorrect affiliation reported in the paper for one of the authors (S. Dulla, Politecnico di Torino). The publisher would like to apologise for any inconvenience caused
A re-visitation of space asymptotic theory in neutron transport
No full textThe space asymptotic theory has constituted a powerful tool for the determination of neutron energy spectra in nuclear reactors, which are the basis of the generation of group constants for the neutronic core design. The method can provide a deep physical insight into the basics of reactor physics and may still give new ideas for modern computational methods. This contribution presents a re-visitation of the method, illustrating its most important general results, some of which may not be well known. In particular, the criticality theory and the space–energy separability theorem are presented. The validity of such theorem is extended also to the net neutron current. The procedure allows to generalize the Fick’s law with a consistent definition of the energy-dependent diffusion coefficient. Some numerical examples are given in simple multigroup models to illustrate the relevant features of the theory
Cross sections polynomial axial expansion within the APOLLO3 3D characteristics method
No full textA polynomial expansion for cross sections has been introduced in the 3D Method Of Characteristics (MOC) solver of the APOLLO3 code, with the aim of applying the method to depleted systems. This paper addresses the suitable adaptation of the MOC equations with regards to the free and DPN-accelerated power iterations, describing also how the conceived algorithms take advantage of a vectorized implementation.
An adaptive quadrature technique is presented, necessary for the evaluation of terms that, due to the polynomial description, cannot be solved analytically. The decisive advantage of the new method over a classical step-constant description of nuclear properties is assessed by comparing the
results of the depletion study of a PWR assembly
Uncertainty quantification in steady state simulations of a molten salt system using polynomial chaos expansion analysis
No full textUncertainty 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 non-intrusive 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.RST/Reactor Physics and Nuclear Material
Safety assessment: perspectives for next generation nuclear plants
No full textSafety assessment and risk analysis are recognized as a priority in the development of next gen-eration nuclear systems (Generation-IV reactors and full-scale fusion reactor –DEMO-) and demand a recon-sideration of the safety philosophy currently applied to the existing nuclear stations. Since their innovative physics and technology and the preliminary design phase of some of the concepts, their safety assessment has to rely on the basis of nuclear safety and technological neutral methodology. In order to satisfy this necessity, a bibliographic survey on nuclear and non-nuclear international standards and best practices is performed. By comparing them, this work tries to reach a new and more systematic approach, based on functional safety, suitable for dealing with the unique challenges of the innovative nuclear facilities, in order to guarantee that safety achievement is intended to be “built-in” rather than “added-on” by influencing the concept evolution from its earliest stages
Generalized perturbation techniques for uncertainty quantification in lead-cooled fast reactors
No full textThe design of innovative nuclear fission systems requires a careful evaluation of the uncertainties affecting the basic input data. Among them, nuclear data are particularly relevant, due to their dramatic energy dependence. Because of this feature and of the strong spatial heterogeneity of nuclear reactors arrangement, full-core calculations are carried out using energy collapsed and spatially homogenised constants. Nowadays, collapsing is often performed with Monte Carlo codes, which allow a discretisation-free treatment of the neutron transport equation. The most popular method to propagate the uncertainty in the nuclear data libraries throughout the Monte Carlo transport calculation is the Generalised Perturbation Theory (GPT). However, due to its multi-group nature, GPT often blurs the continuous-energy feature of the Monte Carlo method. Therefore, in order to fully exploit its advantages, the XGPT method has been recently proposed. After discussing the main differences between these two approaches, the paper presents the application to an uncertainty quantification study on the lead-cooled fast reactor ALFRED design, performed with GPT and focused on the multi-group cross sections. Afterwards, the two nuclides that most contribute to the overall uncertainties, i.e. Pu-239 and U-238, are considered to compare the GPT results to some XGPT calculations carried out with different multi-group energy structures. This analysis suggests that XGPT is a consistent method for uncertainty quantification in the continuous-energy Monte Carlo framework. Moreover, it can be concluded that an adequate number of low-energy groups is necessary for an accurate uncertainty evaluation in the case of a fast system
The time eigenvalue spectrum for nuclear reactors in multi-group diffusion theory
No full textWe develop a fully analytical study of the spectrum of the neutron diffusion operator both for
spatially homogeneous and reflected reactors in a multi-group energy model. We illustrate and discuss the
results of the analysis of the time spectrum of the diffusion operator, to highlight some general properties of
the neutronic evolution in a multiplying system. Various new results are presented, particularly regarding
the possible existence of complex time eigenvalues, the appearance of a continuum part of the spectrum
and the orthogonality properties of the eigenfunctions in the case of an infinite reflect
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