1,724,419 research outputs found
Statistical and Deterministic Analyses on ADS-demo Fuel Rod by using TRANSURANUS code
No full textThe ADS-demo is an intermediate step in the development of the Accelerator Driven System (ADS) aimed at the transmutation of Pu, MAs and LLFPs. This report concerns some aspects the ADS-demo fuel pin modelling in the so-called 'Enhanced Nominal Conditions'-case B (5). Some key points ara discussed by utilising TRANSURANUS code as the thermal feedback, FGR and corrosion behaviour of LBE-AISI316L system. Cladding swelling behaviour triggers thermal teedback and FGR (gas grain boundary concentration limit input is crucial for the ADS fuel pin modelling). The adopted corrosion hypothesis (30% cladding thickness corrosion at EOL) does not modify previous canclusions (5) on design limits veriticatian. The great flexibility and power of TRANSURANUS fuel rod analysis has been newly confirmed by developing the custom code and the statistic-deterministic mixed strategy herein adopted. The report refers to the P894 FIS-NUC project and more specifically to the activity items detailed in the work plan (1)
Sustainability issues of plutonium recycling in light water reactors: Code evaluations up to 2050
No full textPlutonium recycling in light water reactors is a viable and mature technology capable of improving several indicators of great importance in the evaluation of the sustainability of nuclear energy development in the near- and long-term such as the shortage of natural uranium resources, the increase of spent fuel inventories and the proliferation risks of high-level radioactive waste. This paper, after a brief review of the status of plutonium recycling, presents the results of a scenario analysis carried out in the hypothesis that the share of nuclear fleet loaded with mixed uranium-plutonium oxide fuel (MOX) is kept constant up to the middle of the century. Beside mentioned indicators, the paper discusses the needs for complex and costly fuel cycle infrastructures required for the reprocessing of spent nuclear fuel and the fabrication of MOX. Assuming that the deployment of fast reactors occurs beyond the middle of the century, the article focuses on the comparison of an open fuel cycle strategy with a closed fuel cycle strategy where plutonium is recycled. Presented calculations, according to moderate and high projections of nuclear energy development, confirm that plutonium recycling, although deployed to a limited extent, could be beneficial in reducing the stockpiles of nuclear spent fuel and in reducing the risks of proliferation due to the amount of fissile plutonium in the system. The improvement found in the consumption of natural uranium resources was limited promoting to this purpose the deployment of next-generation fast reactors. If a moderate development of nuclear energy is confirmed, the current capacity for reprocessing and MOX fuel fabrication could be sufficient to cope with the foreseen demand, on the contrary, significant investments could be necessary in case of steep increase of installed nuclear energy. Calculations were performed by means of the DESAE code (Dynamic Energy System-Atomic Energy), a tool developed within the IAEA INPRO project. © 2013 Elsevier Ltd. All rights reserved
Thermophysical properties of inert matrix fuels for actinide transmutation
No full textThe thermophysical properties of inert matrix fuels (IMF) for actinide transmutation were discussed. The heat capacity, thermal diffusivity and conductivity were found to be measured in the usable temperature range for MgO- and ZrO2- based IMFs containing uranium, plutonium and americium. The comparison of thermal transport properties of a set of IMFs was also presented
Investigations on the Italian Nuclear Scenario
Full text linkLa recente modifica della politica energetica Italiana è caratterizzata, nel lungo termine, dall'obiettivo di ricavare una quota pari aI 25% del fabbisogno di energia elettrica da fonte nucleare, avendo come termine temporale il 2030. In questo rapporto, si riportano i risultati ottenuti in due scenari, uno di riferimento, l'altro di sviluppo, caratterizzati da una capacità nucleare installata di 19.5 e 35 GW(e) rispettivamente. Utilizzando il codice DESAE (Dynamic Energy System - Atomic Energy), di origine IAEA, sono state investigate le opzioni ciclo del combustibile aperto e riciclo del plutonio con l'intento di studiare la performance di sistemi nucleari ad acqua leggera rispetto a parametri quali consumo di uranio, quantità di combustibile spento, accumulo di plutonio fissile ed attinidi minori. A questo riguardo, i risultati confermano che sistemi nucleari caratterizzati da elevati valori di burnup sono più performanti. Per contro, sistemi a basso burnup sono più attrattivi, sulla base di un maggiore accumulo di plutonio fissile, qualora si pianificasse l'introduzione di reattori veloci e la chiusura del ciclo del combustibile. Il confronto tra ciclo aperto e riciclo del plutonio richiede approfondimenti di carattere sia tecnico che economico
The Heat Capacity of PuO2 at High Temperature: Molecular Dynamics Calculations
No full textA new generation of fast breeder reactors (FBRs) is under development with the objective of making nuclear energy more sustainable. Most promising reactor designs are loaded, at least during their early phase of deployment, with UO2-PuO2 mixed oxide fuel (MOX). Concentrations of plutonium dioxide that are foreseen for FBRs range up to 30 mol%. This highlights the need for a sound and deep knowledge of the thermophysical properties of PuO2. This statement is valid in the case of heat capacity, as evaluations on MOX fuel are usually carried out by using the Neumann-Kopp rule. Heat capacity is relevant for thermal conductivity and performance under transient conditions. However, measurements on the heat capacity of plutonium dioxide are scarce or even lacking at high temperature. Numerical methodologies such as molecular dynamics (MD) calculations have been employed to overcome the difficulties encountered in experimental measurements. Besides numerical also theoretical models have been applied as valuable tools for interpretation of enthalpy measurements. Nevertheless, due to the mentioned lack of experimental measurement issues such as the existence of the Bredig transition and the formation of defects at high temperatures are still debated in nuclear fuel research. Excess enthalpy seen in measurements of actinides oxides has been explained by means of either electronic disorder or anion disorder. In the case of plutonium dioxide, a common consensus has been reached on the hypothesis that anion disorder leads to a significant increase in heat capacity at high temperature. Konings and Beneš have developed a model that accounts for this phenomenon. Their correlation has been often included in models of heat capacity and employed for recommendations. However, in the high-temperature region, MD calculations showed an underestimation of model predictions that was not compensated by the presence of a peak of heat capacity that has been interpreted as the Bredig transition. Based on these observations, this paper presents MD evaluations on the heat capacity of PuO2 at high temperature that are mostly focused on the formation energy of oxygen Frenkel pairs (OFPs) and its correlation with the model proposed by Konings and Beneš. Besides an interatomic potential published in the open literature and developed in compliance with the experimental thermal expansion of PuO2, a second interatomic potential has been applied in calculations. This latter is featured by a lower formation energy of OFP. The contribution due to defects formation was calculated by means of a simplified theoretical model of heat capacity. Results of calculations in the very high-temperature domain showed an increase in the contribution due to OFP defects consistent with the model by Konings and Beneš. Predictions suggest the onset of a premelting transition around 85% of melting temperature without the presence of a peak of heat capacity. Major deviations from the recommended model have been noted in the intermediate temperature region where the effect of clustering of defects should play a significant role. Therefore, the value of formation energy of OFP proposed by Konings and Beneš could be interpreted as an effective value that accounts for the two processes (defects clustering and premelting transition) that could contribute, according to our results, to the heat capacity of plutonium dioxide at high temperature. This conclusion is consistent with the numerical evaluations of OFP formation energy that are in general higher than proposed by Konings and Beneš
Melting temperature of MOX fuel for FBR applications: TRANSURANUS modelling and experimental findings
No full textThe paper is focused on the modelling of the melting temperature of mixed oxide (MOX) fuel for fast breeder reactors (FBRs). After a review of the models available in the TRANSURANUS (TU) code and in the open literature, their predictions were compared to an experimental dataset compiled from published measurements. The recommended model of TRANSURANUS was confirmed to be in good agreement with experimental data. A critical discussion of the comparison provided additional useful indications for the future development of the code and for the recommendations to the users involved in the analysis of the performance of fast reactor fuel. A special attention was given to the presence of minor actinides (MA), a topic of great importance for closure of the nuclear fuel cycle. In this frame, the code could be extended with the model of Konno in order to account for the presence of minor actinides. Finally, the review of the experimental data indicated the need for a reassessment of the effect of the oxygen-to-metal (O/M) ratio on the melting temperature in the low plutonium content domain of relevance to FBR fuel. © 2014 Elsevier B.V. All rights reserved
D67 App. D - Fuel Rod Thermal and Mechanical Analysis
No full textThis report is focused on the PDS-XADS fuel rod thermal and mechanical performance analysis. The Design Basis Conditions - Category I (Normal Operations) were assumed. In particular two cases have been considered: case A (nominal conditions), case B (linear heat rate and fast neutron flux increased by 20% with respect to case A as well as irradiation time increased by 50%). Results confirm that design limits are fully respected far both cases. The Helium pre-pressurisation effect was specifically investigated. The analysis suggests that increasing the fabrication tilling gas pressure would be beneficial for the overall fueI rod performance. Further analysis has been performed to investigate the PDS-XADS fuel rod behaviour when a zero lower plenum volume is assumed. In this case a significant difference is found under conditions B at the End Ot Life, nevertheless design limits are still respected
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