A method for the continuous monitoring of reactivity in subcritical source-driven systems

The reactivity monitoring in subcritical accelerator-driven systems is a key aspect for the development of this technology. In this work, an inverse method for the determination of the system reactivity from the analysis of flux and power signals, based on the point kinetic approach, is applied to source-driven systems. The features of the algorithm specific to the application to subcritical assemblies are identified, and the sensitivity to the integral parameters characterizing the system is discussed. The technique is applied to different transient situations, simulated by neutronic codes adopting point kinetics and multigroup diffusion, and its accuracy in the presence of localized spatial and spectral phenomena is assessed. Different approaches for the reduction of the uncertainties introduced by the experimental noise are proposed and compared

A non-intrusive reduced order model for neutronic transient analyses of the ALFRED reactor

The safety analysis of Gen-IV lead-cooled fast reactors requires accurate evaluations for different operational and accidental conditions. The FRENETIC code, constituted by a full-core, multi-group nodal diffusion module coupled with a thermal-hydraulics one, allows to perform accurate transient calculations, but it is not suitable for the detailed parametric evaluations required by a thorough safety assessment. To fill this gap in the code performances, a non-intrusive reduced-order model reproducing an accurate approx- imation of the FRENETIC output with a reduced computational effort is proposed. The results obtained for a stand-alone neutronic transient involving the accidental insertion of a control rod show how the approach adopted is promising for computationally-efficient safety assessments

“On the Foundation of Transport-Driven Diffusion for Neutron Transport Problems”

The article presents the foundation of a novel methodology developed for the solution of the neutron transport equation, named the transport driven-diffusion approach, which can be considered as an evolution of the classic multiple collision method. The idea behind this method is based on the expansion of the full solution in terms of the contributions of the particles emitted by successive collisions plus a residual term, accounting for particles which have undergone more than a predefined number of collisions. In order to determine the contribution at each collision order, a transport equation with a source term is solved, while the estimation of the residue is based on a diffusion theory model. The physical rationale for the choice of the diffusion model for the residue is discussed and justified, as physics suggests that the diffusion assumptions become more applicable for the description of the particles having suffered a certain number of collisions rather than to the original transport problem. Some results are presented for a set of steady-state and time-dependent test cases. Their analysis shows the remarkable advantage of the method proposed in terms of accuracy and computational time, when compared to standard diffusion and multiple collision at the same order

Application of the Polynomial Chaos Expansion to the Uncertainty Propagation in Fault Transients in Nuclear Fusion Reactors: DTT TF Fast Current Discharge

Nuclear fusion reactors are composed of several complex components whose behavior may be not certain a priori. This uncertainty may have a significant impact on the evolution of fault transients in the machine, causing unexpected damage to its components. For this reason, a suitable method for the uncertainty propagation during those transients is required. The Monte Carlo method would be the reference option, but it is, in most of the cases, not applicable due to the large number of required, repeated simulations. In this context, the Polynomial Chaos Expansion has been considered as a valuable alternative. It allows us to create a surrogate model of the original one in terms of orthogonal polynomials. Then, the uncertainty quantification is performed repeatedly, relying on this much simpler and faster model. Using the fast current discharge in the Divertor Tokamak Test Toroidal Field (DTT TF) coils as a reference scenario, the following method has been applied: the uncertainty on the parameters of the Fast Discharge Unit (FDU) varistor disks is propagated to the simulated electrical and electromagnetic relevant effects. Eventually, two worst-case scenarios are analyzed from a thermal–hydraulic point of view with the 4C code, simulating a fast current discharge as a consequence of a coil quench. It has been demonstrated that the uncertainty on the inputs (varistor parameters) strongly propagates, leading to a wide range of possible scenarios in the case of accidental transients. This result underlines the necessity of taking into account and propagating all possible uncertainties in the design of a fusion reactor according to the Best Estimate Plus Uncertainty approach. The uncertainty propagation from input data to electrical, electromagnetic, and thermal hydraulic results, using surrogate models, is the first of its kind in the field of the modeling of superconducting magnets for nuclear fusion applications

Coupled modelling of the EBR-II SHRT-45R including photon heat deposition

The Fast REactor NEutronics/Thermal-hydraulICs (FRENETIC) code has been developed during the last years at Politecnico di Torino, implementing a full-core coupled neutronic/thermal-hydraulics model for steady-state and transient analysis of liquid-metal cooled fast breeder reactor (LMFBR). In the framework of the validation activities for the code, an analysis of the sodium-cooled reactor EBR-II, previously carried out in the frame of a IAEA Coordinated Research Project, is performed with FRENETIC including the most recent physics models. In particular, photon transport and heat deposition are taken into account, a feature which has been proved in previous studies to be relevant to the correct study of the EBR-II core. To this purpose, a set of nuclear data for photons has been generated by means of the Monte Carlo code Serpent-2, and it is demonstrated that the code is able to take into account the photon heat deposition in the EBR-II

Convergence acceleration aspects in the solution of the PN neutron transport eigenvalue problem

The solution of the eigenvalue problem for neutron transport is of utmost importance in the field of reactor physics, and represents a challenging problem for numerical models. Different eigenvalue formulations can be identified, each with its own physical significance. The numerical solution of these problems by deterministic methods requires the introduction of approximations, such as the spherical harmonics expansion in PN models, leading to results that depend on the approximations introduced (spatial mesh size, N order, ...). All these results represent, in principle, sequences that can easily profit from acceleration techniques to approach convergence towards the correct value. Such a reference value is estimated, in this work, by the Monte Carlo technique. The Wynn- acceleration method is applied to the various sequences of eigenvalues emerging when tackling the solution of the PN models with different orders and increasing spatial accuracy, in order to obtain more accurate, benchmark-quality results. It is shown that the acceleration can be successfully applied and that the analysis of the results of different acceleration approaches sheds some light on the physical meaning of the numerical approximations

Study of an intrinsically safe infrastructure for training and research on nuclear technologies

Within European Partitioning & Transmutation research programs, infrastructures specifically dedicated to the study of fundamental reactor physics and engineering parameters of future fast-neutron-based reactors are very important, being some of these features not available in present zero-power prototypes. This presentation will illustrate the conceptual design of an Accelerator-Driven System with high safety standards, but ample flexibility for measurements. The design assumes as base option a 70MeV, 0.75mA proton cyclotron, as the one which will be installed at the INFN National Laboratory in Legnaro, Italy and a Beryllium target, with Helium gas as core coolant. Safety is guaranteed by limiting the thermal power to 200 kW, with a neutron multiplication coefficient around 0.94, loading the core with fuel containing Uranium enriched at 20% inserted in a solid-lead diffuser. The small decay heat can be passively removed by thermal radiation from the vessel. Such a system could be used to study, among others, some specific aspects of neutron diffusion in lead, beam-core coupling, target cooling and could serve as a training facility