Papers by Guglielmo Lomonaco
EPJ Nuclear Sciences & Technologies, Dec 31, 2022
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Energies
To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental ... more To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect experiments, and more complex phenomena can be investigated using integral-effect tests, the behavior of an entire system as complex as a nuclear power plant core must be investigated using computer code modeling. One of the most commonly used computer codes for the assessment of severe accidents is MELCOR 2.2. In version 18019, the authors enabled the modeling of the behavior of the nuclear fuel with FeCrAl cladding (namely, alloy B136Y3) for the first time, using the GOX model. The ability of this model to reasonably accurately predict the behavior of FeCrAl cladding in accident conditions with quenching was verified in this work by modeling the QUENCH-19 experiment carried out in the Karlsruhe Institute of Technology on the QUENCH experimen...
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Energies, Jun 14, 2024
The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in ... more The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in recent years as an intermediate step before fusion energy. This work proposes a possible approach to FFHRs based on the coupling of a Reversed Field Pinch fusion machine and a Molten Salt Subcritical fission test bed. The proposed test bed is characterized by the coexistence of a fast-neutron fission core and a dedicated thermal-neutron zone, allowing the performing of tritium breeding and actinides transmutation studies. The neutronic design solutions and the results obtained by the irradiation of FLiBe salt (inside the thermal-neutron zone) and of an actinide target (inside the core) are shown. The outcomes of the analysis reveal the potential of FFHR systems as breeding/burner systems. In particular, the results regarding tritium breeding are very encouraging as the system is demonstrated to be able to reach a very high Tritium Breeding Ratio.
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Energies, Jun 14, 2024
The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in ... more The studies on the development of fusion–fission hybrid reactors (FFHR) have gained consensus in recent years as an intermediate step before fusion energy. This work proposes a possible approach to FFHRs based on the coupling of a Reversed Field Pinch fusion machine and a Molten Salt Subcritical fission test bed. The proposed test bed is characterized by the coexistence of a fast-neutron fission core and a dedicated thermal-neutron zone, allowing the performing of tritium breeding and actinides transmutation studies. The neutronic design solutions and the results obtained by the irradiation of FLiBe salt (inside the thermal-neutron zone) and of an actinide target (inside the core) are shown. The outcomes of the analysis reveal the potential of FFHR systems as breeding/burner systems. In particular, the results regarding tritium breeding are very encouraging as the system is demonstrated to be able to reach a very high Tritium Breeding Ratio.
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EPJ Nuclear Science and Technology, Dec 19, 2023
New nuclear technologies are currently being study to face High Level Waste treatment and disposa... more New nuclear technologies are currently being study to face High Level Waste treatment and disposal issues. Generally, GEN-IV fission Fast Reactors (FR) are considered the waste-burners of the future. In fact, a fast flux turns out to be the best choice for actinides irradiation in critical reactors because of favorable cross section conditions. Differently, Fusion Fission Hybrid Reactors (FFHR) are futuristic devices based on the combination of fusion and fission systems and could represent an alternative to FRs. In such systems, the choice spectrum of the neutron flux that irradiates HLW may be non-obvious due to some operational constraints which have to be considered. To design and optimize these systems as waste-burners, one should fully understand the transmutation dynamics occurring into the fission region. A multi-energy-group analysis by FISPACT-II code has been set to analyze the conversion processes in scenarios characterized by different neutron energy spectra and fluences. The results of this study show that, despite fast fluxes are characterized by better behaviors in terms of radiotoxicity treatment, the difficulties of reaching high reaction yields may require solutions involving moderators or broadened neutron fluxes to increase the reactions probabilities and, consequently, actinides mass conversion yield.
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... 8 Page 10. 2 La Sicurezza Nucleare degli impianti esistenti e futuri ... Sovietica) al reat... more ... 8 Page 10. 2 La Sicurezza Nucleare degli impianti esistenti e futuri ... Sovietica) al reattore numero 4, rappresenta il più grave evento nella storia delle applicazioni ... Nei reattori di tipo occidentale infatti l'acqua leggera (H2O) svolge contemporaneamente il ...
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A revised textbook gives an up-to-date overview of the nuclear fuel cycle from mining to radioact... more A revised textbook gives an up-to-date overview of the nuclear fuel cycle from mining to radioactive waste disposal. Aimed at nonspecialists, it looks to be a suitable replacement for some of the older classical texts on this topic
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To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental ... more To ensure the applicability of accident-tolerant fuels, their behaviors under various accidental conditions must be assessed. While the dependences of the behavior of single physical parameters can be investigated in single- or separate-effect experiments, and more complex phenomena can be investigated using integral-effect tests, the behavior of an entire system as complex as a nuclear power plant core must be investigated using computer code modeling. One of the most commonly used computer codes for the assessment of severe accidents is MELCOR 2.2. In version 18019, the authors enabled the modeling of the behavior of the nuclear fuel with FeCrAl cladding (namely, alloy B136Y3) for the first time, using the GOX model. The ability of this model to reasonably accurately predict the behavior of FeCrAl cladding in accident conditions with quenching was verified in this work by modeling the QUENCH-19 experiment carried out in the Karlsruhe Institute of Technology on the QUENCH experimental device and by subsequent comparison of the MELCOR calculation results with the experiment. This article proves that the GOX model can be used to evaluate the behavior of FeCrAl cladding and that the results can be considered conservative.
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Third International Workshop on Technology and Components of Accelerator-Driven Systems (TCADS-3), 2017
We report on the studies of an irradiation facility based on an accelerator-driven subcritical nu... more We report on the studies of an irradiation facility based on an accelerator-driven subcritical nuclear research reactor, which can simultaneously provide a fast flux in the core and a thermal flux in the reflector, that we will call a hybrid fast-slow ADS. The conceptual design presented here, inspired by [1], starts from a 432 kW (keff=0.967) ADS composed by 110 solid lead fuel assemblies each with size 9.7 x 9.7 x 87 cm3, filled with 81 MOX pins (16.5% Pu+Am) of 0.357 cm radius and surrounded by a 0.068 cm thick AISI steel cladding. Source neutrons are produced by a 70 MeV 1 mA proton beam impinging on a beryllium target (~ 8 x 1014 n/sec) [2].The core is cooled by helium flowing in very thin pipes, 0.25 cm in diameter and is surrounded by a 80 cm lead reflector. Core and reflector are contained within a 2 cm steel vessel. The hybrid version (keff 0.972= P=527 kW) is instead composed by 59 fuel assemblies, each hosting 81 MOX pins (22% Pu+Am) where: - The lead reflector has been replaced by three concentric layers, the first of 35 cm lead, followed by 50 cm graphite and finally 10 cm lead. - In the cooling system water flows in wider pipes, 0.5 cm in diameter, which allows to increase keff while maintaining the fast character of the spectrum. We simulated the neutron flux in three core positions (internal, intermediate and external) and in two graphite reflector positions (internal, intermediate), finding that the flux is still mostly fast in the core, while it exhibits a strong thermal component in the reflector, as shown in the following table. This work is partially supported by the 7th Framework Programmes of the European Commission (Euratom) through the CHANDA contract FP7-Fission-2013-605203
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18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019, 2019
In the framework of MYRRHA Project, a pool-type experimental and material testing irradiation fac... more In the framework of MYRRHA Project, a pool-type experimental and material testing irradiation facility operated with Lead Bismuth Eutectic (LBE) coolant and able to operate in both sub-critical and critical mode is designed to be built in Mol, Belgium, in SCK\u2022CEN domain. SCK\u2022CEN entered the pre-licensing phase for the MYRRHA reactor. A complete safety analysis must be performed to provide the necessary data to the safety authority. An event with potential serious consequences is the Primary Heat Exchanger Tube Rupture (PHXTR), involving the release of water from a failed tube in a hot liquid metal pool. In the first phase of a PHXTR accident, the water in the Secondary Cooling System is released in the Primary System pool in choked flow regime. Afterwards, the water bubble formation and characterization is important for the definition of the water specific volume increase and for the estimation of the water mass fraction redirected in the reactor Lower Plenum, with the risk of void insertion in the core and consequent reactivity excursion. An analytical calculation model to evaluate the evolution of any bubble distribution has been set up. The main purpose is to describe the evolution of the main system variables during the accidental event, by checking the potential insurgency of any reactor safety issue due to pressure peaks or core void insertion
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Mathematics
The aim of this work is to present a new analytical model to evaluate jointly the mechanical inte... more The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-...
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Global Journal of Energy Technology Research Updates, 2015
The fusion-fission hybrid reactor is a promising technology that is likely to assume an increasin... more The fusion-fission hybrid reactor is a promising technology that is likely to assume an increasingly important role in the global energy scene in the coming years. This kind of reactor can use both the nuclear fusion and fission processes to produce energy: neutrons from fusion reactions are used to sustain the fission of a sub-critical system. This method allows to have an intrinsically safe facility, with higher efficiency than a fusion reactor itself and with a harder neutron energy spectrum than a fission reactor, which could be suitable for nuclear waste transmutation. This paper, in particular, analyzes a type of hybrid reactor for the transmutation of Minor Actinides (MA). Nuclear waste, in the oxide form, is inserted as an element of the First Wall (FW) of an ITER-like fusion reactor. The aim is to demonstrate the feasibility of the transmutation of the MA characterized by higher long term radiotoxicity into shorter lived nuclides. The neutron transport in a detailed 3D geom...
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The High-Temperature gas-cooled Reactor (HTR) is a promising concept for the next generation of n... more The High-Temperature gas-cooled Reactor (HTR) is a promising concept for the next generation of nuclear power plants, and it is essential that validated analytical tools are available in the European nuclear community to perform conceptual design studies, industrial calculations (reload calculations and the associated core follow), safety analyses for licensing, etc., for new fuel cycles aiming at plutonium and minor actinide (MA) incineration/transmutation without multi-reprocessing of the discharged fuel. The “HTR-N” project of the European Union Fifth Framework Program includes activities concerning the validation of computational tools and the qualification of models. These activities are centred around the two HTR systems currently in operation, viz. the HTR-10 and the HTTR. The re-calculation of the HTTR first criticality with a Monte Carlo neutron transport code now yields acceptable correspondence with experimental data. Also calculations by 3-D diffusion theory codes yield ...
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The fusion-fission hybrid reactor is a promising technology that is likely to assume more and mor... more The fusion-fission hybrid reactor is a promising technology that is likely to assume more and more importance in the global energy scenario in the coming years. Although this kind of nuclear system dates back to the earliest times of the fusion projects (when it was recognized that using fusion neutrons to \u201csupport\u201d nuclear fission fuel cycle could widely increase the exploitation of the fusion plants), it appears to receive relatively limited attention since the mid-1980s. Notwithstanding, hybrid fusion fission systems have been already studied for some decades, in the most prominent laboratories and a relatively large bibliography was produced. Obviously much more papers on this topic have been published in more recent years. The fusion-fission hybrid concept can use both the nuclear fusion and fission processes: in a typical application, neutrons from fusion reactions can be used to sustain the fission chain of a sub-critical system. This is the basis of the hybrid reac...
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Mathematics, May 23, 2022
The aim of this work is to present a new analytical model to evaluate jointly the mechanical inte... more The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works.
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Problems of Atomic Science and Technology, Ser. Thermonuclear Fusion, 2021
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In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in ... more In the context of GEN-IV heavy liquid metal-cooled reactors safety studies, the flow blockage in a fuel sub-assembly is considered one of the main issues to be addressed and one of the most important accidents for Lead Fast Reactors (LFR). In order to model the temperature and velocity field inside a wrapped fuel assembly under unblocked and blocked conditions, a detailed experimental campaign as well as 3D thermal hydraulic analyses of the fuel assembly are required. The present paper is focused on the CFD modeling and preliminary computational analysis of the new experimental facility „Blocked‟ Fuel Pin bundle Simulator (BFPS) that will be inserted in the heavy liquid metal NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The BFPS test section aims to carry out suitable experiments to fully investigate different flow blockage regimes in a 19-pin fuel bundle providing experimental data in support of the development of...
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EPJ Nuclear Sciences & Technologies, 2017
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Papers by Guglielmo Lomonaco
— the RFP physics and the scaling laws at increased level of current and machine size;
— the technologies allowing increased RFP performances and continuous operation;
— a test bench for a hybrid blanket, combining Tritium production and fission reactions.
In order to optimize the pilot experiment approach in terms of cost reduction, best use of the step-by-step acquired knowledge and clear milestones towards the realization of a low power FFHR, a staged approach with increased complexity and investment is introduced [1]. In the proposed pilot FFHR the aim is the production of D—T-fusion power with a RFP configuration (P ≈ 30 MW, Q ≈ 0.4, continuous pulsed operation) and testing the blanket with limited fission fuel. The overall strategy of this approach and the details for each stage of the plant requirements, the tackled issues and the expected results in order to pass to the next phase will be present-ed in the talk.