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Design and Safety Issues of Nuclear Plants and Installations

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A special issue of Energies (ISSN 1996-1073). This special issue belongs to the section "B4: Nuclear Energy".

Deadline for manuscript submissions: closed (31 December 2021) | Viewed by 35412

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Special Issue Editors

Prof. Dr. Rosa Lo Frano

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Guest Editor

Department of Civil and Industrial Engineering, University of Pisa, 56126 Pisa, Italy
Interests: nuclear energy; nuclear technology; safety design; long-term operation; ageing; fusion/fission plants; external event
Special Issues, Collections and Topics in MDPI journals

Dr. Mariano Tarantino

E-Mail Website
Guest Editor

Fusion and Technology for Nuclear Safety and Security Department, Innovative Project Section, ENEA, 40032, Località Brasimone, Camugnano, Bologna, Italy
Interests: nuclear energy; nuclear technolgy; nuclear safety; large scale experiments; thermal-hydraulics; liquid metals technology; LWRs; LFRs; SFRs; fusion reactors

Special Issue Information

Dear Colleagues,

The proposed Special Issue dedicated to design and safety issues and installations focuses on nuclear energy as a sustainable, clean, and safe energy source. It will address the crucial and decisive factors enabling the sustained long-term operation, safety, and profitability of nuclear systems, which can be achieved through a combination of applying optimum management strategies with an understanding of the ways in which safety-relevant systems, structures, and components (SSCs) perform and impact on plant availability, reliability and performance.

This Special Issue will cover nuclear power plant (including SMR) design and operation, including innovative design and SMR, safety assessment as well as related technologies. It will provide a forum to discuss and present recent and innovative research methodologies and results, technologies, experiments and best practices on nuclear power plants, for both fission and fusion technologies. Research results on cascading/conjunct events characterization, fragility analyses and uncertainties treatment will also be included in this Special Issue. Papers can also include research results on advanced and innovative nuclear fuel cycles.

The technologies considered in this Special Issue will enable nuclear systems to guarantee safe operational performances. Moreover, papers that review and implement robust methodologies for assessing SSC fitness-for-service, and identify the most critical elements of the systems that may lesser (or impair) the plant safety margin will be also welcome.

All those and other issues concerning experimental and numerical analysis of nuclear systems will be dealt with. Finally, the proposed Special Issue will also bridge research with educational programs, as well as engineering practices, in all disciplines related to nuclear technology.

It is my pleasure to invite you to submit a manuscript for it. Full papers, communications, and reviews are all welcome.

Prof. Dr. Rosa Lo Frano
Dr. Mariano Tarantino
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. Energies 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 2600 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

fusion/fission plants technology
nuclear fuel and related technologies
nuclear plant engineering
nuclear power plants (including SMR) design and operation
operating plant experience
probabilistic risk assessment
deterministic safety assessment
beyond design basis
nuclear safety

Published Papers (14 papers)

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16 pages, 3636 KiB

Open AccessArticle

Uncertainty and Sensitivity Analysis of a Dry Cask for Spent Nuclear Fuel

by Michela Angelucci, Sandro Paci, Francisco Feria and Luis Enrique Herranz

Energies 2022, 15(3), 1216; https://0-doi-org.brum.beds.ac.uk/10.3390/en15031216 - 07 Feb 2022

Cited by 3 | Viewed by 1601

Abstract

Nuclear safety relies to a good extent on thoroughly validated codes. However, code predictions are affected by uncertainties that need to be quantified for a more accurate evaluation of safety margins. In this regard, the present paper proposes a preliminary uncertainty and sensitivity analysis of the thermal behavior of a concrete-based dry cask for spent nuclear fuel storage, employing the MELCOR code and a series of MATLAB scripts. As thermal behavior is of utmost importance for the fulfillment of United States Nuclear Regulatory Commission (USNRC) safety requirements, the Peak Cladding Temperature (PCT) has been addressed as the key Figure of Merit (FOM). Variables related to the main heat transfer mechanisms have been selected as input parameters for the uncertainty quantification, whereas heat source and heat sink, namely decay power and external air temperature, have been dealt with in a separate sensitivity analysis. The results show that the selected parameters have a weak influence on the PCT, whereas it is strongly related to the decay power and external air temperature values. In any case, PCT stays below the regulatory threshold even under the considered off-normal conditions. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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13 pages, 2794 KiB

Open AccessArticle

MARIA Reactor Irradiation Technology Capabilities towards Advanced Applications

by Marek Migdal, Emilia Balcer, Łukasz Bartosik, Łukasz Bąk, Agnieszka Celińska, Justyna Cybowska, Krzysztof Dobrzelewski, Janusz Jaroszewicz, Krzysztof Jezierski, Natalia Knake, Wojciech Kubiński, Jan Lechniak, Maciej Lipka, Gaweł Madejowski, Adam Małkiewicz, Łukasz Murawski, Ireneusz Owsianko, Bartłomiej Piwowarski, Rafał Prokopowicz, Anna Talarowska, Emilia Wilińska, Tomasz Witkowski, Piotr Witkowski, Grzegorz Wojtania and Marcin Wójcik Show full author list Hide full author list

Energies 2021, 14(23), 8153; https://0-doi-org.brum.beds.ac.uk/10.3390/en14238153 - 05 Dec 2021

Cited by 6 | Viewed by 3045

Abstract

The MARIA research reactor is designed and operated as a multipurpose nuclear installation, combining material testing, neutron beam experiments, and medical and industrial radionuclide production, including molybdenum-99 (⁹⁹Mo). Recently, after fuel conversion to LEU and rejuvenation of the staff while maintaining their experience, MARIA has been used to respond to the increased interest of the scientific community in advanced nuclear power studies, both fission and fusion. In this work, we would like to introduce MARIA’ s capabilities in the irradiation technology field and how it can serve future nuclear research worldwide. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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33 pages, 10261 KiB

Open AccessArticle

Overview on Lead-Cooled Fast Reactor Design and Related Technologies Development in ENEA

by Mariano Tarantino, Massimo Angiolini, Serena Bassini, Sebastiano Cataldo, Chiara Ciantelli, Carlo Cristalli, Alessandro Del Nevo, Ivan Di Piazza, Dario Diamanti, Marica Eboli, Angela Fiore, Giacomo Grasso, Francesco Lodi, Pierdomenico Lorusso, Ranieri Marinari, Daniele Martelli, Francesca Papa, Camillo Sartorio, Marco Utili and Alessandro Venturini

Energies 2021, 14(16), 5157; https://0-doi-org.brum.beds.ac.uk/10.3390/en14165157 - 20 Aug 2021

Cited by 25 | Viewed by 5780

Abstract

The next generation of nuclear energy systems, also known as Generation IV reactors, are being developed to meet the highest targets of safety and reliability, sustainability, economics, proliferation resistance, and physical protection, with improved performances compared with the currently licensed plants or those presently being built. Among the proposed technologies, lead-cooled fast reactors (LFRs) have been identified by nuclear industries in both Western and developing countries as being among the optimal Generation IV candidates. Since 2000, ENEA, the Italian National Agency for New Technologies, Energy, and Sustainable Economic Development is supporting the core design, safety assessment, and technological development of innovative nuclear systems cooled by heavy liquid metals (HLM) and, most recently, fully oriented on LFRs. ENEA is developing world-recognized skills in fast spectrum core design and is one of the largest European fleets of experimental facilities aiming at investigating HLM thermal-hydraulics, coolant chemistry control, corrosion behavior for structural materials, and material properties in the HLM environment, as well as at developing corrosion-protective coatings, components, instrumentation, and innovative systems, supported by experiments and numerical tools. Efforts are also dedicated to develop and validate numerical tools for specific application to HLM systems, ranging from neutronics codes, system and core thermal-hydraulic codes, computational fluid dynamics (CFD) and fuel pin performance codes, including their coupling. The present work aims at highlighting the capabilities and competencies developed by ENEA so far in the framework of liquid metal technologies for Generation IV LFRs. In particular, an overview on the ongoing R&D experimental program will be depicted considering the current fleet of facilities, namely: CIRCE, NACIE-UP, LIFUS5, LECOR, BID-1, HELENA, RACHEL, and Mechanical Labs. An overview on the numerical activities performed so far and those presently ongoing is also reported. Finally, an overview of the ENEA contribution to the ALFRED Project in the frame of the FALCON international consortium is reported, mainly addressing the ongoing activity in terms of core design, technology development, and auxiliary systems design. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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24 pages, 7021 KiB

Open AccessArticle

A Sensitivity Study of Critical Flow Modeling with MELCOR 2.2 Code Based on the Marviken CFT-21 Experiment

by Mateusz Włostowski, Paweł Domitr and Piotr Darnowski

Energies 2021, 14(16), 4985; https://0-doi-org.brum.beds.ac.uk/10.3390/en14164985 - 13 Aug 2021

Cited by 1 | Viewed by 1858

Abstract

The paper presents a study of the critical flow phenomena modeling with MELCOR 2.2.15254 severe accident computer code. The Marviken Critical Flow Test number 21 (CFT-21) experiment was selected as a representative critical flow-focused Separate Effect Test (SET). Various modeling aspects were investigated, including the nodalization, model setup, parameters, and sensitivity coefficients. A local-type sensitivity study was performed to analytically identify the significant parameters and assess their impact on the modeling. A dedicated regression-based approach, using standard deviation, was developed to find the best-fit MELCOR modeling parameters. The primary purpose of this work was to determine the appropriate approach to model critical flow with MELCOR 2.2, investigate the model performance, assess the influence of nodalization choices, identify significant sensitivity parameters, and prepare recommendations with an emphasis on best-estimate modeling. An additional outcome was the benchmark of the recent code version with the Marviken test. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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21 pages, 15281 KiB

Open AccessArticle

Influence of Different Sites on Impact Response of Steel-Plate Concrete Containment against a Large Commercial Aircraft

by Xiuyun Zhu, Jianbo Li, Gao Lin, Rong Pan and Liang Li

Energies 2021, 14(16), 4957; https://0-doi-org.brum.beds.ac.uk/10.3390/en14164957 - 12 Aug 2021

Viewed by 1160

Abstract

This paper aimed at evaluating the influence of different site conditions on the impact response of the structure of nuclear power plants (NPPs) against a large commercial aircraft. The lumped parameter site dynamic model recommended by the code of ASCE 4-98 was used to consider the different homogeneous sites. With respect to the excellent impact resistant performance of steel-plate concrete (SC) structure, the full SC containment is selected as the research object. The impact analysis of the full SC containment against a large commercial aircraft under different site conditions was carried out, based on the force time-history analysis method. The numerical results in terms of the displacement, plastic strain, local concrete damage, and different values of energy were evaluated. The results showed that: (1) For the relatively thin full SC containment, the impact response under the fixed boundary is the largest, while that calculated by other, different sites varies greatly, and there is no consistent rule, the boundary condition which is assumed to be fixed is relatively conservative. (2) For the thicker full SC containment, the displacement response decreased with the increasing of the site shear wave velocity, which is the smallest when the fixed boundary is considered. When the shear wave velocity of the site is large enough, its boundary condition which is assumed to be the fixed constraint is reasonable. (3) For the relatively thin full SC containment, the site damping effect has a significant effect on the structural impact response. Nevertheless, the impact response of the thicker containment is slightly influenced by the site damping effect. (4) For the impact analysis of the structures of NPPs against a large commercial aircraft, it is suggested that both the specific site condition and fixed boundary should be considered. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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28 pages, 6760 KiB

Open AccessArticle

Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

by Piotr Darnowski, Piotr Mazgaj and Mateusz Włostowski

Energies 2021, 14(16), 4884; https://0-doi-org.brum.beds.ac.uk/10.3390/en14164884 - 10 Aug 2021

Cited by 7 | Viewed by 1958

Abstract

In this study, uncertainty and sensitivity analyses were performed with MELCOR 2.2.18 to study the hydrogen generation (figure-of-merit (FoM)) during the in-vessel phase of a severe accident in a light water reactor. The focus of this work was laid on a large generation-III pressurized water reactor (PWR) and a double-ended hot leg (HL) large break loss of coolant accident (LB-LOCA) without a safety injection (SI). The FPT-1 Phebus integral experiment emulating LOCA was studied, where the experiment outcomes were applied for the plant scale modelling. The best estimate calculations were supplemented with an uncertainty analysis (UA) based on 400 input-decks and Latin hypercube sampling (LHS). Additionally, the sensitivity analysis (SA) utilizing the linear regression and linear and rank correlation coefficients was performed. The study was prepared with a new open-source MELCOR sensitivity and uncertainty tool (MelSUA), which was supplemented with this work. The FPT-1 best-estimate model results were within the 10% experimental uncertainty band for the final FoM. It was shown that the hydrogen generation uncertainties in PWR were similar to the FPT-1, with the 95% percentile being covered inside a ~50% band and the 50% percentile inside a ~25% band around the FoM median. Two different power profiles for PWR were compared, indicating its impact on the uncertainty but also on the sensitivity results. Despite a similar setup, different uncertainty parameters impacted FoM, showing the difference between scales but also a significant impact of boundary conditions on the sensitivity analysis. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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20 pages, 9814 KiB

Open AccessArticle

Influence of Vertical Equivalent Damping Ratio on Seismic Isolation Effectiveness of Nuclear Reactor Building

by Xiuyun Zhu, Jianbo Li, Gao Lin and Rong Pan

Energies 2021, 14(15), 4602; https://0-doi-org.brum.beds.ac.uk/10.3390/en14154602 - 29 Jul 2021

Cited by 3 | Viewed by 1423

Abstract

This paper aimed at evaluating the influence of different vertical equivalent damping ratios of a 3-dimensional combined isolation bearing (3D-CIB) as regards seismic response and isolation effectiveness. A comparative study of the seismic response in terms of acceleration floor response spectra (FRS), peak acceleration, displacement response of the nuclear reactor building, and dynamic response of the 3D-CIB was carried out. The results showed that: (1) the horizontal FRS is slightly influenced by the vertical equivalent damping ratio of 3D-CIB, whereas the increase of the vertical equivalent damping ratio has a significant effect on reducing the vertical FRS; (2) the peak vertical acceleration increased with the decrease in the vertical equivalent damping ratios of 3D-CIB and the difference of peak accelerations calculated by the damping ratio of 20 and 25% is within 10%; (3) the increase of the vertical equivalent damping ratio is capable of reducing the horizontal displacement and the rocking effect of the superstructure, and effectively controlling the vertical displacement amplitude; and (4) the vertical equivalent damping ratio of 3D-CIB has a slight effect on its axial force. Consequently, it is demonstrated that the increase of the vertical equivalent damping ratio is advantageous for isolation effectiveness. From the view of displacement control, it is suggested that the 3D-CIB with the vertical an equivalent damping ratio of 15~20% is appropriate and acceptable. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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11 pages, 1440 KiB

Open AccessArticle

The EC MUSA Project on Management and Uncertainty of Severe Accidents: Main Pillars and Status

by Luis Enrique Herranz, Sara Beck, Victor Hugo Sánchez-Espinoza, Fulvio Mascari, Stephan Brumm, Olivia Coindreau and Sandro Paci

Energies 2021, 14(15), 4473; https://0-doi-org.brum.beds.ac.uk/10.3390/en14154473 - 24 Jul 2021

Cited by 22 | Viewed by 2373

Abstract

In the current state of maturity of severe accident codes, the time has come to foster the systematic application of Best Estimate Plus Uncertainties (BEPU) in this domain. The overall objective of the HORIZON-2020 project on “Management and Uncertainties of Severe Accidents (MUSA)” is to quantify the uncertainties of severe accident codes (e.g., ASTEC, MAAP, MELCOR, and AC²) when modeling reactor and spent fuel pools accident scenarios of Gen II and Gen III reactor designs for the prediction of the radiological source term. To do so, different Uncertainty Quantification (UQ) methodologies are to be used for the uncertainty and sensitivity analysis. Innovative AM measures will be considered in performing these UQ analyses, in addition to initial/boundary conditions and model parameters, to assess their impact on the source term prediction. This paper synthesizes the major pillars and the overall structure of the MUSA project, as well as the expectations and the progress made over the first year and a half of operation. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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15 pages, 3451 KiB

Open AccessArticle

Preliminary Analysis of an Aged RPV Subjected to Station Blackout

by Rosa Lo Frano, Salvatore Angelo Cancemi, Piotr Darnowski, Riccardo Ciolini and Sandro Paci

Energies 2021, 14(15), 4394; https://0-doi-org.brum.beds.ac.uk/10.3390/en14154394 - 21 Jul 2021

Cited by 5 | Viewed by 2062

Abstract

Today, 46% of operating Nuclear Power Plants (NPP) have a lifetime between 31 and 40 years, while 19% have been in operation for more than 40 years. Long Term Operation (LTO) is an urgent requirement for all of the nuclear industry. The aim of this study is to assess the performance of a reactor pressure vessel (RPV) subjected to a station blackout (SBO) event. Alterations suffered by the material properties and creep at elevated temperatures are considered. In this study, coupling between MELCOR and Finite Element Method (FEM) codes is carried out. In the Finite Element (FE) model, the combined effects of ageing and creep are implemented through degraded material properties and a viscoplastic model. The reliability of the model is validated by comparing the FOREVER/C1 experimental results. The results show that the RPV lower head bends downwards with a maximum radial expansion of about 260 mm and RPV thermomechanical properties are reduced by more than 50% at high temperatures. The effects of ageing, creep and long heat-up strongly affect the resistance of the RPV system until the point of compromising it in the absence of/delayed emergency intervention. Aged RPV at end-of-life may collapse earlier, and in less time, with the same accidental conditions. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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15 pages, 2598 KiB

Open AccessArticle

Research on Improved Seismic Instrumentation System for Nuclear Power Plants

by Liang Li, Xiuli Du, Rong Pan, Xiuyun Zhu and Haiyan Luan

Energies 2021, 14(14), 4262; https://0-doi-org.brum.beds.ac.uk/10.3390/en14144262 - 14 Jul 2021

Cited by 1 | Viewed by 1495

Abstract

According to the requirements of nuclear safety regulations, nuclear power plants must be equipped with seismic instrumentation systems, which are mainly used for monitoring alarm and automatic shutdown alarm during an earthquake. Both the second and third generation NPPs adopt Peak Ground Acceleration (PGA). However, among the seismic acceleration characteristics, isolated and prominent single high frequency acceleration peaks have no decisive influence on the seismic response. Especially when the earthquake monitoring alarm is at 1 out of 7, it is likely to cause a false alarm or false shutdown. In addition, it usually takes one month or more for the NPPs to restart after the shutdown. In this paper, an improved seismic instrumentation system based on the existing system is proposed. For high intensity areas, three components resultant acceleration is used to judge the 2 out of 4 logic of the automatic seismic trip system(ASTS). For low intensity areas, the seismic failure level is evaluated quickly by using three components resultant acceleration, seismic instrument intensity, cumulative absolute velocity, floor response spectrum and other multi-parameters, avoiding unnecessary and long-term shutdown inspection. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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22 pages, 8872 KiB

Open AccessArticle

Overview of a Theory for Planning Similar Experiments with Different Fluids at Supercritical Pressure

by Andrea Pucciarelli, Sara Kassem and Walter Ambrosini

Energies 2021, 14(12), 3695; https://0-doi-org.brum.beds.ac.uk/10.3390/en14123695 - 21 Jun 2021

Cited by 3 | Viewed by 1481

Abstract

The recent advancements achieved in the development of a fluid-to-fluid similarity theory for heat transfer with fluids at supercritical pressures are summarised. The prime mover for the development of the theory was the interest in the development of Supercritical Water nuclear Reactors (SCWRs) in the frame of research being developed worldwide; however, the theory is general and can be applied to any system involving fluids at a supercritical pressure. The steps involved in the development of the rationale at the basis of the theory are discussed and presented in a synthetic form, highlighting the relevance of the results achieved so far and separately published elsewhere, with the aim to provide a complete overview of the potential involved in the application of the theory. The adopted rationale, completely different from the ones in the previous literature on the subject, was based on a specific definition of similarity, aiming to achieve, as much as possible, similar distributions of enthalpies and fluid densities in a duct containing fluids at a supercritical pressure. This provides sufficient assurance that the complex phenomena governing heat transfer in the addressed conditions, which heavily depend on the changes in fluid density and in other thermophysical properties along and across the flow duct, are represented in sufficient similarity. The developed rationale can be used for planning possible counterpart experiments, with the aid of supporting computational fluid-dynamic (CFD) calculations, and it also clarifies the role of relevant dimensionless numbers in setting up semi-empirical correlations for heat transfer in these difficult conditions, experiencing normal, enhanced and deteriorated regimes. This paper is intended as a contribution to a common reflection on the results achieved so far in view of the assessment of a sufficient body of knowledge and understanding to base successful predictive capabilities for heat transfer with fluids at supercritical pressures. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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20 pages, 9928 KiB

Open AccessArticle

Sensitivity Analysis of Steel-Plate Concrete Containment against a Large Commercial Aircraft

by Xiuyun Zhu, Jianbo Li, Gao Lin, Rong Pan and Liang Li

Energies 2021, 14(10), 2829; https://0-doi-org.brum.beds.ac.uk/10.3390/en14102829 - 14 May 2021

Cited by 3 | Viewed by 1537

Abstract

Due to the excellent impact resistant performance of steel-plate concrete (SC) structure compared with the conventional reinforced concrete (RC) structure, SC structure is preferred to be used in the design of external walls of nuclear island buildings for new nuclear power plants (NPPs). This study aims at evaluating the effect of material and geometric parameters of SC containment on its impact resistant performance, thus the numerical simulation and sensitivity analysis of SC containment subjected to malicious large commercial aircraft attack are conducted based on the force time-history analysis method. The results show that: (1) the impact resistant performance of full SC containment is better than that of half SC containment; (2) for relatively thin full SC containment, the impact response and concrete damage can be significantly reduced by the enhancing of concrete strength grade or the increasing of steel plate thickness; (3) for the thicker full SC containment, concrete strength grade has only a slight influence on the impact displacement response, and the increasing of steel plate thickness has no significant effect on mitigating the impact displacement response. However, the increasing of steel plate thickness can effectively reduce its plastic strain, and the decreasing of strength grade of steel plate may obviously increase its plastic strain; and (4) concrete thickness plays a decisive role on the improvement of impact resistance, which is more effective than the enhancing of concrete strength grade. Resultantly, this paper provides a reference and guidance for the design of SC structure external walls of nuclear island buildings against a large commercial aircraft. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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Review

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48 pages, 2114 KiB

Open AccessReview

Materials for Sustainable Nuclear Energy: A European Strategic Research and Innovation Agenda for All Reactor Generations

by Lorenzo Malerba, Abderrahim Al Mazouzi, Marjorie Bertolus, Marco Cologna, Pål Efsing, Adrian Jianu, Petri Kinnunen, Karl-Fredrik Nilsson, Madalina Rabung and Mariano Tarantino

Energies 2022, 15(5), 1845; https://0-doi-org.brum.beds.ac.uk/10.3390/en15051845 - 02 Mar 2022

Cited by 14 | Viewed by 5468

Abstract

Nuclear energy is presently the single major low-carbon electricity source in Europe and is overall expected to maintain (perhaps eventually even increase) its current installed power from now to 2045. Long-term operation (LTO) is a reality in essentially all nuclear European countries, even when planning to phase out. New builds are planned. Moreover, several European countries, including non-nuclear or phasing out ones, have interests in next generation nuclear systems. In this framework, materials and material science play a crucial role towards safer, more efficient, more economical and overall more sustainable nuclear energy. This paper proposes a research agenda that combines modern digital technologies with materials science practices to pursue a change of paradigm that promotes innovation, equally serving the different nuclear energy interests and positions throughout Europe. This paper chooses to overview structural and fuel materials used in current generation reactors, as well as their wider spectrum for next generation reactors, summarising the relevant issues. Next, it describes the materials science approaches that are common to any nuclear materials (including classes that are not addressed here, such as concrete, polymers and functional materials), identifying for each of them a research agenda goal. It is concluded that among these goals are the development of structured materials qualification test-beds and materials acceleration platforms (MAPs) for materials that operate under harsh conditions. Another goal is the development of multi-parameter-based approaches for materials health monitoring based on different non-destructive examination and testing (NDE&T) techniques. Hybrid models that suitably combine physics-based and data-driven approaches for materials behaviour prediction can valuably support these developments, together with the creation and population of a centralised, “smart” database for nuclear materials. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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Other

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14 pages, 2102 KiB

Open AccessProject Report

The H2020 McSAFER Project: Main Goals, Technical Work Program, and Status

by Victor Hugo Sanchez-Espinoza, Stephan Gabriel, Heikki Suikkanen, Joonas Telkkä, Ville Valtavirta, Marek Bencik, Sören Kliem, Cesar Queral, Anthime Farda, Florian Abéguilé, Paul Smith, Paul Van Uffelen, Luca Ammirabile, Marcus Seidl, Christophe Schneidesch, Dmitry Grishchenko and Hector Lestani

Energies 2021, 14(19), 6348; https://0-doi-org.brum.beds.ac.uk/10.3390/en14196348 - 04 Oct 2021

Cited by 14 | Viewed by 2708

Abstract

This paper describes the main objectives, technical content, and status of the H2020 project entitled “High-performance advanced methods and experimental investigations for the safety evaluation of generic Small Modular Reactors (McSAFER)”. The main pillars of this project are the combination of safety-relevant thermal hydraulic experiments and numerical simulations of different approaches for safety evaluations of light water-cooled Small Modular Reactors (SMR). It describes the goals, the consortium, and the involved thermal hydraulic test facilities, e.g., the COSMOS-H (KIT), HWAT (KTH), and MOTEL (LUT), including the experimental programs. It also outlines the different safety assessment methodologies applied to four different SMR-designs, namely the CAREM (CNEA), SMART (KAERI), F-SMR (CEA), and NuScale. These methodologies are multiscale thermal hydraulics, conventional, low order, and high fidelity neutron physical methods used to demonstrate the inherent safety features of SMR-core designs under postulated design-basis-accident conditions. Finally, the status of the investigations is shortly discussed followed by the dissemination activities and an outlook. Full article

(This article belongs to the Special Issue Design and Safety Issues of Nuclear Plants and Installations)

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