1,721,045 research outputs found
A review of the scientific contributions by Massimo Salvatores to nuclear reactor physics and engineering
No full textMassimo Salvatores has been an eminent personality of the nuclear science community. He is universally recognized as one of the scientists who helped to maintain research excellence in the nuclear field throughout the past 50 years. His activity covers the theoretical aspects of nuclear reactor physics, the development of efficient computational methods and tools, and experimental analysis. This paper reviews the scientific production of Massimo Salvatores, highlighting what are believed to be his most relevant contributions to nuclear science and engineering
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
Generalized perturbation techniques for uncertainty quantification in lead-cooled fast reactors
Full text linkThe 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
Reduced order models in reactor kinetics: A comparison between point kinetics and multipoint kinetics
No full textDue to the heavy computational burden of full reactor kinetics modeling, reduced-order models are usually employed to simulate transients. Among those, point and multipoint kinetics have small computation time and provide satisfying results for many applications. We implemented point and multipoint kinetics (Avery's and Kobayashi's models) in the APOLLO3 code. Then these models are applied to study two simple transients, one in a coupled fast-thermal configuration, the other in a fast reactor. As at present we focus on the neutronic response only (neutron and precursor populations), the study is limited to step-change transients with no thermal feedback. This work permits to better delineate the potential of these methods and opens interesting perspectives. As an example, multipoint kinetics allows accounting for very fast shape transients that, depending on the coupling of the system, may result in global population changes occurring before the conventional prompt jump and altering significantly its quantitative value
The adjoint problem as physical heuristic for loading pattern optimization
No full textLoading pattern optimization enables longer irradiation cycles while still maintaining the ability to safely operate and shutdown the reactor. The loading pattern optimization problem is multi-objective and is characterized by a huge nonlinear discrete search space. Its complexity grows exponentially with the number of fuel assemblies in the core. However, in-core fuel management is an essential part of routine work in a nuclear reactor, hence optimized usage of the fuel inventory enables the economic and efficient utilization of resources. This study is a proof-of-concept for the hypothesis that adjoint-based neutron importance functions can be used for the optimization process of the core loading pattern. New optimiza- tion techniques are developed in order to demonstrate the successful utilization of adjoint-based func- tions as the optimization driving force. Different importance functions are developed and studied. It is demonstrated that the physical insight obtained from the importance function can be used for the opti- mization of loading patterns. Eventually, this new technique should be integrated into a stochastic opti- mization algorithm, e.g., evolutionary algorithms, in order to accelerate and improve existing optimization algorithms
Verification and validation of the modular ray tracing MOC using the coupled forward-adjoint approach and application to C5G7 benchmark
No full textNeutron transport adjoint calculations are useful in many reactor physics applications. Among various applications, the adjoint flux can be used in perturbation theory to prevent large calculations and computational costs when estimating a small reactivity insertion into the system. Furthermore, they can serve as validation tests for numerical schemes, since both direct and adjoint calculations for a given system should lead to the same eigenvalue, although using two different physicomathematical formulations of the transport model. The synchronized implementation of the method of characteristic (MOC) for neutron transport in forward and adjoint approaches is accomplished in this work. The result is validated using the C5G7 benchmark with comparisons of the multiplication factor and pin power values. Differences between forward and adjoint multiplication factors in the results are achieved in the order of 1.0E-6. Meanwhile, the difference between the multiplication factor and the C5G7 benchmark is in order of 1.0E-5, using an S16 level symmetric angular discretization and a track spacing of 0.01 cm
On the boundary conditions for the neutron transport equation
No full textThe solution of the linear transport equation used for the study of neutral particle fields requires the imposition of appropriate boundary conditions. The choice of the conditions to impose for an infinite medium is not straightforward. The question has been given different formulations in the literature with various justifications based on some physical reasoning. Some aspects of the question are here analysed, from both the mathematical and the physical point of view. It is concluded that the inspiring golden rule should be the establishment of conditions that do not require any reference to the properties of the specific medium being considered for their justification
On the relation between spherical harmonics and simplified spherical harmonics methods
No full textThe purpose of the paper is, first, to recall the proof that the AN method and, therefore, the SP2N−1 method (of which AN was shown to be a variant) are equivalent to the odd order P2N−1, at least for a particular class of multi-region problems; namely the problems for which the total cross section has the same value for all the regions and the scattering is supposed to be isotropic. By virtue of the introduction of quadrature formulas representing first collision probabilities, this class is then enlarged in order to encompass the systems in which the regions may have different total cross sections. Some examples are reported to numerically validate the procedure
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