1,720,964 research outputs found
Modellazione del comportamento di elio in combustibili MOX ad alto burn-up ai fini dell’estensione degli strumenti di simulazione e di supporto alla progettazione delle barrette di combustibile per i reattori di generazione evolutiva ad acqua pressurizzata
No full textTechnical Report RdS/2011/35
Modelli di rilascio dei gas di fissione per combustibili MOX ad elevato burnup
No full textTechnical Report RdS/2013/02
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
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
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
Applying Advanced Neutron Transport Calculation for Improving Fuel Performance Codes
No full textWe present refinements of the Helium production model implemented in the TRANSURANUS fuel performance code. Helium is produced in oxide fuels by three main paths: (i) alpha decay of the actinides; (ii) (n,a) reactions; and (iii) ternary fission. In this work, the contributions due to ternary fission and the 16O(n,a)13C reaction as well as some refinements in
the 241Am burn-up chain have been included in TRANSURANUS. The 16O(n,a)13C cross section has been evaluated in PWR conditions by means of the MCNP Monte Carlo code for different fuel compositions. The Monte Carlo depletion code VESTA has been used for the validation of the Helium production model. For specific PWR conditions the comparison of TRANSURANUS
predictions with those of VESTA is satisfactory, and the applied cross section library is the main source of uncertainty
A Preliminary Assessment of the T91 Behaviour for the Extension of the TRANSURANUS Code to the Fuel Rod Performance Analysis of LBE Cooled Nuclear Reactors
No full textThis work presents the status quo of the efforts carried out in the last years at the Politecnico di Milano in order to assess the information available in the open literature on the T91 steel, which is designed to be the cladding material in LBE (Lead–Bismuth Eutectic) accelerator-driven systems (ADS), like the Italian XADS and the Belgian MYRRHA, as well as in lead-cooled nuclear reactors (e.g., ELSY – the European Lead System). The work, which is oriented to embody this knowledge in the TRANSURANUS code database for cladding materials, is organized as follows.
In the first part, a critical review of the empirical correlations for the heat transfer to liquid metals is presented, with the aim of identifying the most suitable Nusselt number correlation for the evaluation of the cladding temperature, which plays an important role in the fuel rod performance (e.g., in terms of corrosion, swelling and creep behaviour).
Corrosion resistance being a key issue in the development of LBE/lead-cooled reactors, the well-known Tedmon’s model is discussed in the second part. This model has been employed as the starting point for the prediction of the T91 corrosion behaviour in a LBE or lead environment for the conditions of the so-called “in situ” oxide layer protection: on the basis of experimental data from the recent literature, a preliminary correlation for the T91 corrosion in flowing LBE under oxygen control has been set-up. In the last years, efforts have been spent by several research groups to find ways (e.g., via aluminization) in order to form and keep protective films on the T91 steel, but tests simulating core operational conditions are needed to verify the integrity and the adhesion of such coatings, which would require a specific modelling by means of a fuel performance code like TRANSURANUS.
The third part is focused on the out-of-pile properties of the T91 steel, with the aim of choosing the most reliable correlations to be implemented in TRANSURANUS. Showing a critical comparison, a database is eventually proposed that includes the following thermo-mechanical properties: melting temperature, thermal expansion, specific heat, density, thermal conductivity, yield strength, Young’s modulus, Poisson’s ratio, rupture strain.
In the last part, some relevant phenomena affecting the thermo-mechanical behaviour of the "fuel rod system" under irradiation (like swelling and creep of the cladding steel) are discussed, pointing out some open issues to be solved and the missing experimental data needed for a suitable modelling of the performance in reactor of the T91 steel, so that this new nuclear material can be properly treated by means of the TRANSURANUS code
A Preliminary Approach to the Extension of the TRANSURANUS Code to the Fuel Rod Performance Analysis of HLM-Cooled Nuclear Reactors
No full textThis paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors
Extension of the TRANSURANUS Code to the Analysis of Cladding Materials for Liquid Metal Cooled Fast Reactors: a Preliminary Approach
No full textTechnical Report CESNEF-IN-11-200
Analysing Operational Transients by means of the TRANSURANUS Code
No full textBased on the on-line measurements carried out at Halden as well as on data available from the Super-Ramp project included in the OECD/NEA-IAEA International Fuel Performance Experiments(IFPE) database, a critical analysis is made of the fuel rod behaviour during operational transients by means of the TRANSURANUS fuel performance code. More precisely, recent model changes proposed for the transient gas release and swelling, along with suggestions for an improved simulation of pellet-cladding interaction, are being discussed
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