1,720,995 research outputs found
Modelling of Pellet Cladding Interaction during Power Ramps in PWR Rods by means of TRANSURANUS Fuel Rod Analysis Code
No full textPellet cladding interaction (PCI) in PWR type rods subjected to power ramps was analysed by means of Transuranus (TU) fuel rod performance code. PCI phenomena depend on the fuel power history – i.e. by several irradiation and thermal induced phenomena occurring in the fuel rod and mutually interacting during its life in reactor – and may become critical for cladding integrity under accidental conditions. Ten test fuel rods, whose power histories and post irradiation experiment (PIE) data were available from the OECD/NEA-IAEA International Fuel Performance Experiment (IFPE) Data Base through the Studsvik SUPER-RAMP Project, were simulated by Transuranus. During a power ramp pellet gaseous swelling can be inhibited by cladding pressure and can be over-predicted by a normal operation swelling model. This phenomenon was simulated by a new formulation of a fuel swelling model already available in the code, in order to consider hot pressing of inter-granular fuel porosity due to the high hydrostatic stress resulting from PCI: it was found that Transuranus, as a result of the proposed swelling formulation as well as of the accurate modelling of the other phenomena occurring during irradiation, gives correct predictions on PCI induced fuel rod failures. In addition, PCI failure threshold identified by Transuranus was compared with the technological limits known in literature: the possibility of relaxing these limits for low burn-up values and the preponderance of the European fuel rod design in front of PCI emerged from TU analyses. Finally, a good agreement was found between TU evaluations and PIE data, with regard to fission gas release, fuel grain growth, and creep, corrosion and elongation of the cladding
Analysis of Coupled Dynamics of Molten Salt Reactors
No full textThis paper presents a preliminary analysis of the coupled thermo-hydrodynamics and neutronics of circulating nuclear fuel systems like the thermal Molten Salt Reactor (MSR), one of the "Generation IV International Forum" concepts. This kind of nuclear reactor adopts a molten salt mixture, which flows up through channels in a graphite moderated core and plays the role of both heat generator and coolant. A strongly coupled modelling is needed since the velocity pattern is affected by the neutron dynamics via the heat source from fission reactions, whereas the neutronic behaviour is affected by the thermo-hydrodynamics via the motion of precursors. In this complex environment, featured by a highly non linear regime and a wide disparity of time scales, COMSOL Multiphysics® has been successfully employed to investigate the system behaviour both in steady state and transient conditions with reference to a simple 2-D geometry, which represents a typical channel of a sub-critical MSR and comprises both the flowing fluid and the graphite matrix
Analysis of Thermal-Hydraulic Behaviour of the Molten Salt Nuclear Fuel
No full textThis paper presents a preliminary approach to thermo-hydraulics of the molten salt, which plays the role of both heat generator and coolant in the Molten Salt Reactor (MSR). This kind of nuclear reactor represents one of the "Generation IV International Forum" concepts that can be used for actinides burning, production of electricity, production of hydrogen, and breeding of nuclear fuel. Physics of circulating nuclear fuels, as the molten salt, is featured by a strong coupling between neutronics and thermo-hydrodynamics. In the present study, analyses are performed assuming that the neutronic term is decoupled from fluid dynamics and appears like an energy source term, and taking into account the thermodynamic and transport properties of the molten salt as well as its local flow conditions and heat transfer. Even if this assumption simplifies the equations to be solved, the thermo-hydrodynamic behaviour of the molten salt remains complex. The graphite-moderated channel type molten salt breeder reactor based on a previous research at Oak Ridge National Laboratory (ORNL) is considered: a preliminary study of the heat transfer and pressure losses in a typical MSR core channel is proposed referring to a simple axial-symmetric cylindrical geometry with the aim to investigate the specific behaviour of such system as well as to test and compare two different commercial computer codes – namely, COMSOL® (Multiphysics finite elements software) and FLUENT® (Computational Fluid Dynamics finite volumes software) – on the basis of an analytic framework, in view of their adoption for more realistic, design-oriented and multi-physics simulations
A Generalized Approach to Heat Transfer in Pipe Flow with Internal Heat Generation
No full textIn the last few years, there has been a renewed interest in the Molten Salt Reactor (MSR), one of the "Generation IV International Forum" concepts, which adopts a circulating molten salt mixture as both heat generator (fuel) and coolant. The heat transfer of a fluid with internal heat generation depends on the strength of the source whose influence on the heat exchange process is significant enough to demand consideration. At present, few studies have been performed on the subject from either an experimental or a numerical point of view.
This study considers fluids with a wide range of Reynolds numbers, flowing through smooth and straight circular tubes within which the flow is hydrodynamically developed but thermally developing (conditions of interest for MSR core channels). The study aims at an assessment of the heat transfer modelling for a large variety of fluids (with Prandtl numbers in the range 0 ≤ Pr ≤ 10^4), in particular taking into account the influence of the internal heat generation on the temperature distribution, which plays an important role in the case of molten salts for nuclear reactors. To this purpose, the general and unified solution of the heat transfer equation is applied to the turbulent Graetz problem with boundary conditions of the third kind and arbitrary heat source distribution, incorporating recent formulations for turbulent flow and convection.
Computed results are shown to be in a good agreement with experimental data concerning heat transfer evaluations for both fully developed and thermally developing flow conditions, over a large range of Prandtl numbers (10^-2 < Pr < 10^4). Finally, a preliminary correlation, which includes the Prandtl number range of interest for molten salts, is proposed for the Nusselt number predictions in the case of simultaneous uniform wall heat flux and internal heat generation
A Preliminary Approach to the Neutronics of the Molten Salt Reactor by means of COMSOL Multiphysics
No full textThe Molten Salt Reactor (MSR), proposed along with other five innovative concepts of fission nuclear reactor by the Generation IV International Forum (GIF-IV), represents a challenging task from the modelling perspective because of the strong coupling
between neutronics and thermo-hydrodynamics due to liquid fuel circulation in the primary loop. In this paper COMSOL Multiphysics® is adopted to investigate the MSR neutronics, focusing on the steady-state core average conditions of the
Molten Salt Breeder Reactor (MSBR) developed at Oak Ridge National Laboratory (ORNL). The results achieved by COMSOL, adopting a two energy group diffusion model and using group
constants calculated by means of the deterministic code SCALE5.1, are compared with those achieved by the stochastic code MCNP for validation purpose. In particular, neutron flux profiles and integral quantities, like the effective multiplication
factor and homogenized cross sections, are evaluated and discussed. The model implemented in COMSOL is then used to study the effect of the fuel velocity on the neutronic behaviour of
the analysed MSBR core channel
Physics-based modelling of fission gas swelling and release in UO2 applied to integral fuel rod analysis
No full textA physics-based model is developed for analysing the coupled phenomena of fission gas swelling and release in UO2 fuel during irradiation. The model is featured by a level of complexity suitable for application to integral fuel rod analysis and consistent with the uncertainties pertaining to some parameters. The emphasis is on the modelling of the grain-face gas bubble development and the related dependence of the fission gas swelling and release on the local hydrostatic stress, which is of special importance for the analysis of the fuel behaviour during power ramps and pellet-cladding mechanical interaction conditions. The applicability of the new model to integral fuel rod analysis is verified through implementation
and testing in the TRANSURANUS fuel rod analysis code. In the frame of the IAEA co-ordinated research project on Fuel Modelling at Extended Burn-up FUMEX-III, the model is applied to the simulation
of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database. The comparison of the results with the available experimental data of fission gas swelling and release at moderate burn-up is presented as a first step of validation, pointing out an encouraging predictive accuracy for different irradiation conditions, without any fitting applied to the model parameters
Assessment of the Prediction Capability of the TRANSURANUS Fuel Performance Code on the Basis of Power Ramp Tested LWR Fuel Rods
No full textThe present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to
all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas
release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the "ITU model" for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code
Simulation of Power Ramp Tested LWR Fuel Rods by means of the TRANSURANUS Code
No full textIn the present work, the power histories of 26 LWR fuel rods, which were ramp tested in the Studsvik reactor R2 after base irradiation to burn-ups in the range of 10–45 MWd/kgU, are
simulated by means of the TRANSURANUS fuel performance code. A critical comparison between the code predictions in terms of fission gas release and the experimental data available from
the PWR Super-Ramp and BWR Inter-Ramp Projects is performed. This analysis points out that a systematic under-estimation occurs when the input variables considered as standard TRANSURANUS options are chosen. Since the fuel gaseous swelling under Pellet-Cladding Interaction (PCI) conditions is a current open issue, a preliminary study concerning the modelling of this phenomenon is carried out in order to address future research activities in the framework of the IAEA
Coordinated Research Project FUMEX-III. Influence of different fuel densification and pelletfragment relocation models on fission gas release predictions is also discussed
Extension of the TRANSURANUS Code to the Fuel Rod Performance Analysis of LBE-Cooled Nuclear Reactors
No full textThis work intends to be a starting point for the extension of the TRANSURANUS fuel rod performance code to the modelling of the T91 steel, which is designed to be the cladding material in LBE (lead–bismuth eutectic) accelerator-driven systems (ADS). On the basis of the experimental data available in the recent literature on LBE and T91, a preliminary modelling of the T91 corrosion with flowing LBE under oxygen control is proposed, and the main issues (i.e., heat transfer, creep, swelling) relevant for the performance of this steel in a reactor are discussed, in order to be properly considered in TRANSURANUS
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