A new model with Serpent for the first criticality benchmarks of the TRIGA Mark II reactor

We present a new model, developed with the Serpent Monte Carlo code, for neutronics simulation of the TRIGA Mark II reactor of Pavia (Italy). The complete 3D geometry of the reactor core is implemented with high accuracy and detail, exploiting all the available information about geometry and materials. The Serpent model of the reactor is validated in the fresh fuel configuration, through a benchmark analysis of the first criticality experiments and control rods calibrations. The accuracy of simulations in reproducing the reactivity difference between the low power (10 W) and full power (250 kW) reactor condition is also tested. Finally, a direct comparison between Serpent and MCNP simulations of the same reactor configurations is presented

Object-Oriented Modeling and simulation of a TRIGA reactor plant with Dymola

This work presents the modeling and simulation of a TRIGA-Mark II pool-type reactor with Zirchonium-Hydryde and Uranium fuel immersed in light water, with Modelica object-oriented language, in Dymola simulation environment. The model encompasses the integrated plant system including the reactor pool and cooling circuits. The reactor pool plays a fundamental role in the system dynamics, through a thermal feedback effect on the reactor core neutronics. The pool model is tested against three experimental transients: simulation results are in good accordance with experimental data and provide useful information about the inertial effect of the water inventory on the reactor cooling

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

An improved zero-dimensional model for simulation of TRIGA Mark II dynamic response

The primary aim of this work is to improve the analysis of the dynamic behaviour of the TRIGA Mark II reactor at the University of Pavia through a zero-dimensional approach. Besides the coupling between neutronics (point-reactor kinetics with six delayed neutron precursors group) and thermal-hydraulics (two-region model, with fuel and coolant) implemented in earlier works, the new model considers also the time behaviour of the mass flow rate due to natural circulation, of the neutron poisons and of the primary and secondary pool temperature. The system of coupled first-order differential equations is non-linear, as some state variables, such as the mass flow rate and the coolant temperature, multiply each other. The Simulink™ programming environment for dynamic analysis and control purposes is used to solve the system. A comparison with experimental data collected on-site for different reactor power transients and with measurements of the poison anti-reactivity during reactor shut-down and of the pool temperature allows the validation of the model. The model results and the experimental data reach a remarkable agreement. In addition, a linear stability analysis of the reactor is performed through the root locus and the stability map in terms of the thermal feedback coefficients. This analysis shows how the power level influences the dynamic of the system, and that, for certain values (always negative) of the fuel thermal feedback coefficient, positive values of the one for the moderator still ensures the system stability

A new diagnostic instrument to detect generalized roughness in rolling bearings for induction motors

In this paper the external stray flux around a motor has been investigated as a possible new diagnostic instrument to detect generalized roughness. This very common bearing fault has been slightly considered in the literature so far. Stray flux was measured by two probes, one commercially available and one realized in laboratory; the diagnostic content of both signals has been compared also with the measurement of the stator current. The experimental tests have been carried out on an induction motor by substituting one of its bearings with an artificially damaged one, according to three steps of progressive wear. The three measured variables have been analyzed in the frequency domain, by evaluating in particular the characteristic frequencies of the single-point bearing defects. The diagnostic content of the stray flux has been discovered as definitely higher with respect to the stator current and therefore its use to detect generalized roughness is promising

L.E.N.A. - Laboratory of Applied Nuclear Energy (University of Pavia) - Neutron Facilities & Main Activities

The TRIGA (Training Research and Isotopes-production General Atomics) Mark II nuclear research reactor installed at the Laboratory of Applied Nuclear Energy (L.E.N.A.) of the University of Pavia is licensed for operating at 250 kW power in steady state. Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields offering different in- and out-core neutron irradiation channels. A subcritical assembly is also available at the University of Pavia (Chemistry Department Radiochemistry Area) for experimental activities. Current scientific and Educational &Training activities are also described

Characterization of the TRIGA Mark II reactor full-power steady state

In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics

Characterization of the TRIGA Mark II reactor full-power steady state

In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the available experimental data as benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core is necessary. To evaluate it, a thermal-hydraulic model has been developed, using the power distribution results from MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then introduced in the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configuration. The good agreement between experimental data and simulation results concerning full-power reactor criticality, proves the reliability of the adopted methodology of analysis, both from neutronics and thermal-hydraulics perspective.Comment: 22 pages, 17 figures, 2 appendixes. Submitted to Nuclear Engineering and Design International Journa

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