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Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles
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Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles

A special issue of Sustainability (ISSN 2071-1050). This special issue belongs to the section "Sustainable Engineering and Science".

Deadline for manuscript submissions: closed (31 March 2022) | Viewed by 15729

Special Issue Editors

Prof. Dr. Guglielmo Lomonaco

E-Mail Website
Guest Editor
Department of Mechanical, Energy, Management and Transport Engineering (DIME), University of Genoa, Via all'Opera Pia, 15A, 16145 Genova, GE, Italy
Interests: nuclear energy; nuclear technology; innovative nuclear fuel cycles; neutronics; CFD; advanced nuclear systems; energy scenarios; nuclear hydrogen production; HTR; LFR; GFR; ADS; SMR; nuclear space reactors
Special Issues, Collections and Topics in MDPI journals
Prof. Dr. Antonio Cammi

E-Mail Website
Guest Editor
Department of Energy, Nuclear Engineering Division (CeSNEF) - Politecnico di Milano, via La Masa, 34-20156 Milano, Italy
Interests: nuclear reactor dynamics and control; multi-physics modelling; generation IV nuclear reactors object-oriented simulators; reduced-order modelling approach; thermal hydraulics

Special Issue Information

Dear Colleagues,

This Special Issue of Sustainability focuses on nuclear energy as a sustainable energy source. It will cover nuclear power plants design and operation, as well as related technologies. It will provide forum to discuss and present recent research results, technologies, and best practices on nuclear power plants, for both fission and fusion technologies. Papers can include small modular reactors designs, technologies, and operation. Research results on advanced and innovative nuclear fuel cycles will also be included in this Special Issue.

The proposed Special Issue will address the role of nuclear energy in a future electricity markets that may feature a high presence of renewables and other variable sources as well as inexpensive peaking capacity. The technologies considered in this Special Issue will enable nuclear power systems to realize optimal performances in a sustainable way. In fact one additional purpose of this Special Issue will be to review and debate the sustainability dimensions of nuclear power, based on the latest understanding of nuclear plant technologies, economic costs, climate change and other externalities. Papers that examine multiple dimensions of sustainability (with or without the use of sustainability metrics) will be also welcome.

Finally, the proposed Special Issue will also bridge research with educational programs, as well as engineering practices, in all disciplines related to nuclear technology.

Prof. Guglielmo Lomonaco
Prof. Antonio Cammi
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Sustainability is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2400 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • Advanced Nuclear Systems (including Space Nuclear Reactors)
  • Nuclear Fuel Cycles
  • Nuclear Fusion and Fission Technologies
  • Nuclear Power Plants (including SMR) Design and Operation
  • Nuclear Power Plants Economics
  • Nuclear Power Plants Integration with the Grid
  • Nuclear Reactor Engineering
  • Nuclear Reactor Physics
  • Nuclear Safety and Security
  • Radiation Detection and Protection Systems
  • Sustainability of Nuclear Energy Systems

Published Papers (5 papers)

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Research

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32 pages, 19935 KiB  
Article
Analysis of Release Model Effect in the Transport of Fission Products Simulating the FPT3 Test Using MELCOR 2.1 and MELCOR 2.2
by Alain Flores y Flores, Danilo Ferretto, Tereza Marková and Guido Mazzini
Sustainability 2021, 13(14), 7964; https://0-doi-org.brum.beds.ac.uk/10.3390/su13147964 - 16 Jul 2021
Cited by 1 | Viewed by 1990
Abstract
The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the [...] Read more.
The severe accident integral codes such as Methods for Estimation of Leakages and Consequences of Releases (MELCOR) are complex tools used to simulate and analyse the progression of a severe accident from the onset of the accident up to the release from the containment. For this reason, these tools are developed in order to simulate different phenomena coupling models which can simulate simultaneously the ThermoHydraulic (TH), the physics and the chemistry. In order to evaluate the performance in the prediction of those complicated phenomena, several experimental facilities were built in Europe and all around the world. One of these facilities is the PHEBUS built by Institut de Radioprotection et de Sûrete Nucléaire (IRSN) in Cadarache. The facility reproduces the severe accident phenomena for a pressurized water reactor (PWR) on a volumetric scale of 1:5000. This paper aims to continue the assessment of the MELCOR code from version 2.1 up to version 2.2 underlying the difference in the fission product transport. The assessment of severe accident is an important step to the sustainability of the nuclear energy production in this period where the old nuclear power plants are more than the new reactors. The analyses presented in this paper focuses on models assessment with attention on the influence of B4C oxidation on the release and transport of fission products. Such phenomenon is a concern point in the nuclear industry, as was highlighted during the Fukushima Daiichi accident. Simulation of the source term is a key point to evaluate the severe accident hazard along with other safety aspects. Full article
(This article belongs to the Special Issue Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles)
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17 pages, 892 KiB  
Article
Preliminary Analysis and Design of the Energy Conversion System for the Molten Salt Fast Reactor
by Andrea Di Ronco, Francesca Giacobbo, Guglielmo Lomonaco, Stefano Lorenzi, Xiang Wang and Antonio Cammi
Sustainability 2020, 12(24), 10497; https://0-doi-org.brum.beds.ac.uk/10.3390/su122410497 - 15 Dec 2020
Cited by 6 | Viewed by 2059
Abstract
The unique design features of the molten salt fast reactor (MSFR) should enable higher coolant temperatures than in conventional water reactors, with a significant improvement in the achievable thermodynamic performance. The use of a molten salt as both fuel and coolant, however, poses [...] Read more.
The unique design features of the molten salt fast reactor (MSFR) should enable higher coolant temperatures than in conventional water reactors, with a significant improvement in the achievable thermodynamic performance. The use of a molten salt as both fuel and coolant, however, poses several advanced heat transfer challenges, such as the design of innovative heat exchangers and energy conversion systems. In this work, we address a preliminary but quantitative analysis of the energy conversion system for the MSFR, based on reference design data from the SAMOFAR H2020-EURATOM project. We consider three main technologies, i.e., the supercritical steam cycle, the closed helium cycle and the helium/steam combined cycle. Preliminary design results are presented for each technology, based on a simplified modelling approach. The considered cycles show promising efficiency improvements, with the best performance being proven by the supercritical steam cycle. The analysis also highlights the critical issue related to the risk of freezing of the molten salts within the secondary heat exchangers, due to the low inlet temperatures of the working fluids. Results show potential incompatibility between the freezing point of molten salts and the temperatures typical of steam cycles, while helium cycles offer the best chances of freezing avoidance. The combined cycle promises intermediate performance in terms of thermodynamic efficiency and thermal compatibility with molten salts comparable with closed helium cycles. Full article
(This article belongs to the Special Issue Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles)
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19 pages, 1407 KiB  
Article
A Kalman Filter-Based Approach for Online Source-Term Estimation in Accidental Radioactive Dispersion Events
by Andrea Di Ronco, Francesca Giacobbo and Antonio Cammi
Sustainability 2020, 12(23), 10003; https://0-doi-org.brum.beds.ac.uk/10.3390/su122310003 - 30 Nov 2020
Cited by 2 | Viewed by 2760
Abstract
In the present work, a online data assimilation approach, based on the Kalman filter algorithm, is proposed for the source term reconstruction in accidental events with dispersion of radioactive agents in air. For this purpose a Gaussian plume model of dispersion in air [...] Read more.
In the present work, a online data assimilation approach, based on the Kalman filter algorithm, is proposed for the source term reconstruction in accidental events with dispersion of radioactive agents in air. For this purpose a Gaussian plume model of dispersion in air is embedded in the Kalman filter algorithm to estimate unknown scenario parameters, such as the coordinates and the intensity of the source, on the basis of measurements collected by a mobile sensor. The approach was tested against pseudo-experimental data produced with both the Gaussian plume model and the Lagrangian puff model SCIPUFF. The results show the good capabilities of the proposed approach in retrieving the values of the unknown parameters when (i) one or more release parameters are poorly known and (ii) a sufficient number of experimental measurements describing the evolution of the dispersion process can be collected in a short time by means of mobile sensors. Thanks to its flexibility and computational efficiency, and due to the exploitation of the Kalman filter potentialities through the use of a simplified model of dispersion in air, the proposed approach can constitute a useful tool for the management of emergency scenarios. Full article
(This article belongs to the Special Issue Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles)
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13 pages, 3119 KiB  
Article
Development of Analytical Procedures for Chemical Characterization of Substrates for the Production of TRISO Coated Particles as Nuclear Fuel in High Temperature Gas-Cooled Reactors
by Ewelina Chajduk, Paweł Kalbarczyk, Jakub Dudek, Marta Pyszynska, Anna Bojanowska-Czajka and Zbigniew Samczyński
Sustainability 2020, 12(17), 7221; https://0-doi-org.brum.beds.ac.uk/10.3390/su12177221 - 03 Sep 2020
Cited by 4 | Viewed by 2350
Abstract
High temperature gas-cooled reactors have recently gained importance as a source of electricity and process heat. Nuclear fuel used in these reactors consists of TRISO (TRiple coated ISOtropic) coated particles, where spherical grains of UO2 or UC2 or UCO kernel are [...] Read more.
High temperature gas-cooled reactors have recently gained importance as a source of electricity and process heat. Nuclear fuel used in these reactors consists of TRISO (TRiple coated ISOtropic) coated particles, where spherical grains of UO2 or UC2 or UCO kernel are covered with four successive layers consisting of pyrolytic carbon and silicon carbide. Of great importance is the chemical purity of reagents and substances used for the production of TRISO coated fuel particles. Analytical techniques ensuring the determination of elements at trace levels are inductively coupled plasma mass spectrometry (ICP-MS) and neutron activation analysis (NAA). They were applied in this work for the chemical characterization of substrates used for TRISO fuel production. Two analytical procedures were developed: the first, where materials are analyzed using ICP-MS, and the second with the aid of NAA. Successive stages of these procedures are described with details. Results of quantitative chemical analysis of examined substances are reported as well as detection limits for the investigated elements. Moreover, the expanded uncertainties estimated for the determined elements while employing the devised analytical procedures are presented. Full article
(This article belongs to the Special Issue Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles)
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Review

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37 pages, 4646 KiB  
Review
Lattice Boltzmann Method Applied to Nuclear Reactors—A Systematic Literature Review
by Johan Augusto Bocanegra Cifuentes, Davide Borelli, Antonio Cammi, Guglielmo Lomonaco and Mario Misale
Sustainability 2020, 12(18), 7835; https://0-doi-org.brum.beds.ac.uk/10.3390/su12187835 - 22 Sep 2020
Cited by 8 | Viewed by 5119
Abstract
Nuclear engineering requires computationally efficient methods to simulate different components and systems of plants. The Lattice Boltzmann Method (LBM), a numerical method with a mesoscopic approach to Computational Fluid Dynamic (CFD) derived from the Boltzmann equation and the Maxwell–Boltzmann distribution, can be an [...] Read more.
Nuclear engineering requires computationally efficient methods to simulate different components and systems of plants. The Lattice Boltzmann Method (LBM), a numerical method with a mesoscopic approach to Computational Fluid Dynamic (CFD) derived from the Boltzmann equation and the Maxwell–Boltzmann distribution, can be an adequate option. The purpose of this paper is to present a review of the recent applications of the Lattice Boltzmann Method in nuclear engineering research. A systematic literature review using three databases (Web of Science, Scopus, and ScienceDirect) was done, and the items found were categorized by the main research topics into computational fluid dynamics and neutronic applications. The features of the problem addressed, the characteristics of the numerical method, and some relevant conclusions of each study are resumed and presented. A total of 45 items (25 for computational fluid dynamics applications and 20 for neutronics) was found on a wide range of nuclear engineering problems, including thermal flow, turbulence mixing of coolant, sedimentation of impurities, neutron transport, criticality problem, and other relevant issues. The LBM results in being a flexible numerical method capable of integrating multiphysics and hybrid schemes, and is efficient for the inner parallelization of the algorithm that brings a widely applicable tool in nuclear engineering problems. Interest in the LBM applications in this field has been increasing and evolving from early stages to a mature form, as this review shows. Full article
(This article belongs to the Special Issue Sustainability in Innovative Nuclear Fission/Fusion Systems and Related Fuel Cycles)
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