Feature Paper Special Issue for Editorial Board Members (EBMs) of Journal of Nuclear Engineering

A special issue of Journal of Nuclear Engineering (ISSN 2673-4362).

Deadline for manuscript submissions: closed (31 December 2023) | Viewed by 36995

Special Issue Editor


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Guest Editor
Department of Mechanical Engineering, University of South Carolina, Columbia, SC 29201, USA
Interests: all areas of nuclear engineering; sensitivity and uncertainty analysis of large-scale systems; predictive modeling by combining experimental and computational information to reduce uncertainties in predicted results
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Special Issue Information

Dear Colleagues,

To celebrate the successful launch of the Journal of Nuclear Engineering (JNE), we are devoting this Special Issue of JNE to publishing a selection of articles contributed by the Editorial Board Members (EBMs). This Special Issue aims to highlight various research accomplishments, as well as open issues, in the areas represented by the EBMs. All EBMs are invited to contribute original research articles or critical reviews, which will all be peer reviewed before acceptance for publication.

Prof. Dr. Dan Gabriel Cacuci
Guest Editor

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. Journal of Nuclear Engineering is an international peer-reviewed open access quarterly 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 1000 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

  • reactor statics
  • reactor dynamics
  • reactor safety
  • fuel cycles and nuclide transmutation
  • advanced reactor systems
  • sensitivity analysis
  • uncertainty quantification
  • data assimilation and predictive modeling

Published Papers (14 papers)

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Research

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15 pages, 3464 KiB  
Article
Reliability Assessment of NPP Safety Class Equipment Considering the Manufacturing Quality Assurance Process
by Mohammad Khalaquzzaman, Seung Jun Lee and Muhammed Mufazzal Hossen
J. Nucl. Eng. 2023, 4(2), 421-435; https://doi.org/10.3390/jne4020030 - 2 Jun 2023
Viewed by 1751
Abstract
Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design [...] Read more.
Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design might occur at different stages of manufacturing for various reasons, such as a lack of skilled manpower, deviation of materials, human errors, malfunction of equipment, violation of manufacturing procedure, etc. These deviations can be assessed cautiously and taken into consideration in the final safety analysis report (FSAR) before issuing an operating license. In this paper, we propose a Bayesian belief network for quality assessment of safety class equipment of NPPs with a few examples. The proposed procedure is a holistic approach for estimation of equipment failure probability considering manufacturing deviations and errors. Case studies for safety-class dry transformers and reactor pressurizers employing the proposed method are also presented in this article. This study provides insights for probabilistic safety assessment engineers and nuclear plant regulators for improved assessment of NPP safety. Full article
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16 pages, 28787 KiB  
Article
Bulk Tungsten Fiber-Reinforced Tungsten (Wf/W) Composites Using Yarn-Based Textile Preforms
by Alexander Lau, Jan Willem Coenen, Daniel Schwalenberg, Yiran Mao, Till Höschen, Johann Riesch, Leonard Raumann, Michael Treitz, Hanns Gietl, Alexis Terra, Beatrix Göhts, Christian Linsmeier, Katharina Theis-Bröhl and Jesus Gonzalez-Julian
J. Nucl. Eng. 2023, 4(2), 375-390; https://doi.org/10.3390/jne4020027 - 4 May 2023
Cited by 4 | Viewed by 2310
Abstract
The use of tungsten fiber-reinforced tungsten composites (Wf/W) has been demonstrated to significantly enhance the mechanical properties of tungsten (W) by incorporating W-fibers into the W-matrix. However, prior research has been restricted by the usage of single fiber-based textile fabrics, consisting [...] Read more.
The use of tungsten fiber-reinforced tungsten composites (Wf/W) has been demonstrated to significantly enhance the mechanical properties of tungsten (W) by incorporating W-fibers into the W-matrix. However, prior research has been restricted by the usage of single fiber-based textile fabrics, consisting of 150 µm warp and 50 µm weft filaments, with limited homogeneity, reproducibility, and mechanical properties in bulk structures due to the rigidity of the 150 µm W-fibers. To overcome this limitation, two novel textile preforms were developed utilizing radial braided W-yarns with 7 core and 16 sleeve filaments (R.B. 16 + 7), with a diameter of 25 µm each, as the warp material. In this study, bulk composites of two different fabric types were produced via a layer-by-layer CVD process, utilizing single 50 µm filaments (type 1) and R.B. 16 + 7 yarns (type 2) as weft materials. The produced composites were sectioned into KLST-type specimens based on DIN EN ISO 179-1:2000 using electrical discharge machining (EDM) and subjected to three-point bending tests. Both composites demonstrated enhanced mechanical properties with pseudo-ductile behavior at room temperature and withstood over 10,000 load cycles between 50–90% of their respective maximum load without sample fracture in three-point cyclic loading tests. Furthermore, a novel approach to predict the fatigue behavior of the material under cyclic loading was developed based on the high reproducibility of the composites produced, especially for the composite based on type 1. This approach provides a new benchmark for upscaling endeavors and may enable a better prediction of the service life of the produced components made of Wf/W in the future. In comparison, the composite based on fabric type 1 demonstrated superior results in manufacturing performance and mechanical properties. With a high relative average density (>97%), a high fiber volume fraction (14–17%), and a very homogeneous fiber distribution in the CVD-W matrix, type 1 shows a promising option to be further tested in high heat flux tests and to be potentially used as an alternative to currently used materials for the most stressed components of nuclear fusion reactors or other potential application fields such as concentrated solar power (CSP), aircraft turbines, the steel industry, quantum computing, or welding tools. Type 2 composites have a higher layer spacing compared to type 1, resulting in gaps within the matrix and less homogeneous material properties. While type 2 composites have demonstrated a notable enhancement over 150 µm fiber-based composites, they are not viable for industrial scale-up unlike type 1 composites. Full article
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11 pages, 2186 KiB  
Article
Double Pulse LIBS Analysis of Metallic Coatings of Fusionistic Interest: Depth Profiling and Semi-Quantitative Elemental Composition by Applying the Calibration Free Technique
by Salvatore Almaviva, Francesco Colao, Ivano Menicucci and Marco Pistilli
J. Nucl. Eng. 2023, 4(1), 193-203; https://doi.org/10.3390/jne4010015 - 7 Feb 2023
Viewed by 1460
Abstract
In this work we report the characterization of thin metallic coatings of interest for nuclear fusion technology through the ns double-pulse LIBS technique. The coatings, composed of a tungsten (W) or tungsten-tantalum (W-Ta) mixture were enriched with deuterium (D), to simulate plasma-facing materials [...] Read more.
In this work we report the characterization of thin metallic coatings of interest for nuclear fusion technology through the ns double-pulse LIBS technique. The coatings, composed of a tungsten (W) or tungsten-tantalum (W-Ta) mixture were enriched with deuterium (D), to simulate plasma-facing materials (PFMs) or components (PFCs) of the next generation devices contaminated with nuclear fuel in the divertor area of the vacuum vessel (VV), with special attention to ITER, whose divertor will be made of W. The double pulse LIBS technique allowed for the detection of D and Ta at low concentrations, with a single laser shot and an average ablation rate of about 110 nm. The calibration free (CF-LIBS) procedure provided a semi-quantitative estimation of the retained deuterium in the coatings, without the need of reference samples. The presented results demonstrate that LIBS is an eligible diagnostic tool to characterize PFCs with high sensitivity and accuracy, being minimally destructive on the samples, without PFCs manipulation. The CF-LIBS procedure can be used for the search for any other materials in the VV without any preliminary reference samples. Full article
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15 pages, 7617 KiB  
Article
Enhanced Mechanical Properties of CLAM by Zirconium Alloying and Thermo-Mechanical Processing
by Dongping Zhan, Jihang Li, Dongwei Wang, Huishu Zhang, Guoxing Qiu and Yongkun Yang
J. Nucl. Eng. 2023, 4(1), 127-141; https://doi.org/10.3390/jne4010009 - 17 Jan 2023
Viewed by 1306
Abstract
In this study, we present the effects of 0.004~0.098 wt% Zr and thermo-mechanical processing (TMP) on the microstructure and mechanical properties of the China RAFM steel, CLAM, as a feasibility study for improving mechanical properties. The inclusions in ingots were characterized using optical [...] Read more.
In this study, we present the effects of 0.004~0.098 wt% Zr and thermo-mechanical processing (TMP) on the microstructure and mechanical properties of the China RAFM steel, CLAM, as a feasibility study for improving mechanical properties. The inclusions in ingots were characterized using optical microscope (OM) and scanning electron microscope (SEM), which could be classified as fine simple particles and large complex particles. The complexity of the alloy’s inclusion composition increases with the increasing Zr concentration. The higher the Zr content, the more complex the composition of inclusions in the alloy. The average diameter of inclusions in 0.004Zr steel was the smallest, which was 0.79 μm and the volume fraction was 0.018%. The highest yield strength, tensile strength, elongation, and impact energy of 0.004Zr alloy at room temperature were 548.3 MPa, 679.4 MPa, 25.7%, and 253.9 J. The structure of the TMPed steels was all tempered martensite. With the increase in tempering temperature, the yield and tensile strength of the experimental steel gradually decreased, while the elongation and impact energy gradually increased. The 0.004ZrD and 0.004ZrH alloys had the best yield strength and impact energy, which were 597.9 and 611.8 MPa and 225.9 and 243.3 J, respectively. In addition, the alloys showed good thermal stability during the aging at 600 °C for 1500 h. It was discovered that TMP is a simple and practical industrial technique that could successfully enhance the mechanical properties of CLAM steel without sacrificing impact toughness. Full article
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18 pages, 1931 KiB  
Article
Verification and Validation of the SPL Module of the Deterministic Code AZNHEX through the Neutronics Benchmark of the CEFR Start-Up Tests
by Guillermo Muñoz-Peña, Juan Galicia-Aragon, Roberto Lopez-Solis, Armando Gomez-Torres and Edmundo del Valle-Gallegos
J. Nucl. Eng. 2023, 4(1), 59-76; https://doi.org/10.3390/jne4010005 - 27 Dec 2022
Cited by 1 | Viewed by 1714
Abstract
A new module for the AZtlan Nodal HEXagonal (AZNHEX) code, which is part of the AZTLAN Platform, was recently developed based on the Simplified Spherical Harmonics (SPL) scheme to deal with the challenges presented in small fast reactor cores, [...] Read more.
A new module for the AZtlan Nodal HEXagonal (AZNHEX) code, which is part of the AZTLAN Platform, was recently developed based on the Simplified Spherical Harmonics (SPL) scheme to deal with the challenges presented in small fast reactor cores, such as the China Experimental Fast Reactor (CEFR), with high leakage and significant scattering effects. For the verification and validation process, we generated nodal homogenized macroscopic cross-sections (XS) through a full heterogeneous core model using the stochastic code SERPENT and subsequently, these XS were employed in AZNHEX. To verify the SPL implementation, several mesh sensitivity exercises were performed demonstrating that the SPL module was implemented successfully. Furthermore, to validate the code with this new implementation, we modeled some exercises contained in the CEFR benchmark with AZNHEX and compared the results with the experimental data available. The final results show a great improvement compared with the original diffusion solver reducing the deviations significantly from experimental data. In conclusion, it is shown and discussed the relevance of improved numerical models (transport approximations instead of diffusion) for the deterministic calculations of small fast reactors. Full article
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11 pages, 1117 KiB  
Article
The Contribution of Small Modular Reactors to the Resilience of Power Supply
by Francesco Di Maio, Lorenzo Bani and Enrico Zio
J. Nucl. Eng. 2022, 3(2), 152-162; https://doi.org/10.3390/jne3020009 - 24 May 2022
Cited by 5 | Viewed by 3669
Abstract
In recent years, there has been a growing interest in the design, development and commercialization of nuclear power Small Modular Reactors (SMRs). Actual SMR designs cover the full spectrum of nuclear reactor technologies, including water-, gas-, liquid-metal-, and molten-salt-cooled. Despite physical and technological [...] Read more.
In recent years, there has been a growing interest in the design, development and commercialization of nuclear power Small Modular Reactors (SMRs). Actual SMR designs cover the full spectrum of nuclear reactor technologies, including water-, gas-, liquid-metal-, and molten-salt-cooled. Despite physical and technological differences, SMRs share some relevant design features, such as small size, modularity, inherent and passive safety systems. These features are expected to enhance availability, recoverability, promptness and robustness, thereby contributing to the resilience of power supply. Thanks to the peculiar design features of SMRs, they are likely to satisfy a number of Functional Requirements (FRs) for this objective, namely: (i) low vulnerability to external hazards; (ii) natural circulation of primary coolant; (iii) prompt, unlimited and independent core cooling under shutdown conditions; (iv) shutdown avoidance in response to variations of the offsite power supply quality and electrical load; (v) island mode operation; (vi) robust load-following; (vii) independent, self-cranking start. These make advanced Nuclear Power Plants (aNPPs) comprised of SMRs perfect candidates to withstand a broader range of natural disruptions and to recover faster from them, compared to conventional Nuclear Power Plants (cNPPs), thus rendering them a major potential asset for guaranteeing resilience and security of power supply. The review focuses on Natural Technological (NaTech) events that impact a typical Integrated Energy System (IESs) within which SMRs are embedded: IESs are, indeed, being developed to integrate different power generation plants with gas facilities, through gas and electricity infrastructures, because they are expected to bring increased security and resilience of power supply, as shown in the qualitative case study presented. Full article
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33 pages, 393 KiB  
Article
Fourth-Order Comprehensive Adjoint Sensitivity Analysis Methodology for Nonlinear Systems (4th-CASAM-N): II. Application to a Nonlinear Heat Conduction Paradigm Model
by Dan Gabriel Cacuci
J. Nucl. Eng. 2022, 3(1), 72-104; https://doi.org/10.3390/jne3010005 - 24 Feb 2022
Cited by 1 | Viewed by 2032
Abstract
This work illustrates the application of the fourth-order comprehensive sensitivity analysis methodology for nonlinear systems (abbreviated as “4th-CASAM-N”), which enables the efficient computation of exactly determined 1st-, 2nd-, 3rd-, and 4th-order functional derivatives of results produced by computational models with respect to the [...] Read more.
This work illustrates the application of the fourth-order comprehensive sensitivity analysis methodology for nonlinear systems (abbreviated as “4th-CASAM-N”), which enables the efficient computation of exactly determined 1st-, 2nd-, 3rd-, and 4th-order functional derivatives of results produced by computational models with respect to the model’s parameters. Results produced by computational models are called model “responses” and the respective functional derivatives are called “sensitivities” (with respect) to model parameters. The qualifier “comprehensive” indicates that the 4th-CASAM-N methodology enables the exact and efficient computation not only of response sensitivities with respect to customary model parameters (including computational input data, correlations, initial and/or boundary conditions) but also with respect to imprecisely known material boundaries, as would be caused by manufacturing tolerances. The 4th-CASAM-N enables the hitherto very difficult, if not intractable, exact computation of all of the 1st-, 2nd-, 3rd-, and 4th-order response sensitivities for large-scale systems involving many parameters, as usually encountered in practice. A paradigm model that describes nonlinear heat conduction through a material has been chosen to illustrate the application of the 4th-CASAM-N methodology, as this model enables the derivation of tractable closed-form analytical expressions of representative 1st-, 2nd-, 3rd-, and 4th-order response sensitivities while largely avoiding side-tracking algebraic manipulations. The responses chosen for this paradigm model include not only physically measurable quantities but also a synthetic response designed to illustrate the enormous possible reduction in the number of computation when using the 4th-CASAM-N (rather than other methods) for computing response sensitivities. Full article
35 pages, 445 KiB  
Article
Fourth-Order Comprehensive Adjoint Sensitivity Analysis Methodology for Nonlinear Systems (4th-CASAM-N): I. Mathematical Framework
by Dan Gabriel Cacuci
J. Nucl. Eng. 2022, 3(1), 37-71; https://doi.org/10.3390/jne3010004 - 23 Feb 2022
Cited by 3 | Viewed by 2059
Abstract
This work presents the fourth-order comprehensive sensitivity analysis methodology for nonlinear systems (abbreviated as “4th-CASAM-N”) for exactly and efficiently computing the first-, second-, third-, and fourth-order functional derivatives (customarily called “sensitivities”) of physical system responses (i.e., “system performance parameters”) to the system’s (or [...] Read more.
This work presents the fourth-order comprehensive sensitivity analysis methodology for nonlinear systems (abbreviated as “4th-CASAM-N”) for exactly and efficiently computing the first-, second-, third-, and fourth-order functional derivatives (customarily called “sensitivities”) of physical system responses (i.e., “system performance parameters”) to the system’s (or model) parameters. The qualifier “comprehensive” indicates that the 4th-CASAM-N methodology enables the exact and efficient computation not only of response sensitivities with respect to the customary model parameters (including computational input data, correlations, initial and/or boundary conditions) but also with respect to imprecisely known material boundaries, caused by manufacturing tolerances, of the system under consideration. The 4th-CASAM-N methodology presented in this work enables the hitherto very difficult, if not intractable, exact computation of all of the first-, second-, third-, and fourth-order response sensitivities for large-scale systems involving many parameters, as usually encountered in practice. Notably, the implementation of the 4th-CASAM-N requires very little additional effort beyond the construction of the adjoint sensitivity system needed for computing the first-order sensitivities. The application of the principles underlying the 4th-CASAM-N to an illustrative paradigm nonlinear heat conduction model will be presented in an accompanying work. Full article
32 pages, 7214 KiB  
Article
A Rate Theory Model of Radiation-Induced Swelling in an Austenitic Stainless Steel
by Malcolm Griffiths, Juan Ramos-Nervi and Larry Greenwood
J. Nucl. Eng. 2021, 2(4), 484-515; https://doi.org/10.3390/jne2040034 - 23 Nov 2021
Cited by 6 | Viewed by 3156
Abstract
Many rate theory models of cavity (void) swelling have been published over the past 50 years, all having the same, or similar, structures. A rigorous validation of the models has not been possible because of the dearth of information concerning the microstructures that [...] Read more.
Many rate theory models of cavity (void) swelling have been published over the past 50 years, all having the same, or similar, structures. A rigorous validation of the models has not been possible because of the dearth of information concerning the microstructures that correspond with the swelling data. Whereas the lack of microstructure information is still an issue for historical swelling data, in the past 10–20 years data have been published on the evolution of the microstructure (point defect yields from collision cascades, cavity number densities, and dislocation densities/yield strengths) allowing certain gaps in information to be filled when considering historic swelling data. With reasonable estimates of key microstructure parameters, a standard rate theory model can be applied, and the model parameter space explored, in connection with historical swelling data. By using published data on: (i) yield strength as a function of dose and temperature (to establish an empirical expression for dislocation density evolution); (ii) cavity number densities as a function of temperature; and (iii) freely migrating defect (FMD) production as a function of primary knock-on atom (PKA) spectrum, the necessary parameter and microstructure inputs that were previously unknown can be used in model development. This paper describes a rate-theory model for void swelling of 316 stainless steel irradiated in the EBR-2 reactor as a function of irradiation temperature and neutron dose. Full article
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10 pages, 1076 KiB  
Article
A Study of the Minimum Thermal Power of a Nuclear Reactor
by Keith E. Holbert
J. Nucl. Eng. 2021, 2(4), 412-421; https://doi.org/10.3390/jne2040031 - 20 Oct 2021
Viewed by 3030
Abstract
The minimum mass for a critical reactor is well studied whereas the minimum heat production from a nuclear reactor has received little attention. The thermal power of a (sub)critical reactor originates from fission as well as radioactive decay. Fission includes neutron-induced and spontaneous [...] Read more.
The minimum mass for a critical reactor is well studied whereas the minimum heat production from a nuclear reactor has received little attention. The thermal power of a (sub)critical reactor originates from fission as well as radioactive decay. Fission includes neutron-induced and spontaneous fission. For an idealized critical core, we find that the minimum theoretical power is ER/Λ, whereas for a subcritical reactor comprising fissionable material undergoing spontaneous fission, the minimum power is dictated by subcritical multiplication. Interestingly, radioisotopic heat generation exceeds the minimum theoretical fission power for most of the fissile materials examined in this study. Full article
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18 pages, 5362 KiB  
Article
Neutronic Characteristics of ENDF/B-VIII.0 Compared to ENDF/B-VII.1 for Light-Water Reactor Analysis
by Kang-Seog Kim and William A. Wieselquist
J. Nucl. Eng. 2021, 2(4), 318-335; https://doi.org/10.3390/jne2040026 - 23 Sep 2021
Cited by 4 | Viewed by 3328
Abstract
The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion [...] Read more.
The Evaluated Nuclear Data File (ENDF)/B-VIII.0 data library was released in 2018. To assess the new data during development and shortly after release, many validation calculations were performed with static, room-temperature benchmarks. Recently, when performing validation of ENDF/B-VIII.0 for pressurized water reactor depletion calculations, a regression in performance compared to ENDF/B-VII.1 was observed. This paper documents extensive benchmark calculations for light-water reactors performed using continuous energy Monte Carlo code with ENDF/B-VII.1 and -VIII.0 and neutronic characteristics of ENDF/B-VIII.0 are discussed and compared to those of ENDF/B-VII.1. It is our recommendation that ENDF/B data library assessment should include reactor-specific benchmark assessments, including depletion cases, such that these types of regressions may be caught earlier in the data development cycle. Full article
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Review

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37 pages, 8578 KiB  
Review
Strain Localisation and Fracture of Nuclear Reactor Core Materials
by Malcolm Griffiths
J. Nucl. Eng. 2023, 4(2), 338-374; https://doi.org/10.3390/jne4020026 - 4 May 2023
Cited by 5 | Viewed by 1976
Abstract
The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer [...] Read more.
The production of prismatic dislocation loops in nuclear reactor core materials results in hardening because the loops impede dislocation motion. Yielding often occurs by a localised clearing of the loops through interactions with gliding dislocations called channeling. The cleared channels represent a softer material within which most of the subsequent deformation is localized. Channeling is often associated with hypothetical dislocation pileup and intergranular cracking in reactor components although the channels themselves do not amplify stress as one would expect from a pileup. The channels are often similar in appearance to twins leading to the possibility that twins are sometimes mistakenly identified as channels. Neither twins nor dislocation channels, which are bulk shears, produce the same stress conditions as a pileup on a single plane. At high doses, when cavities are produced (either He-stabilised bubbles at low temperatures or voids at high temperatures), there can be reduced ductility because the material is already in an equivalent advanced stage of microscopic necking. He-stabilised cavities form preferentially on grain boundaries and at precipitate or incoherent twin/ε-martensite interfaces. The higher planar density of the cavities, coupled with the incompatibility at the interface, results in a preferential failure known as He embrittlement. Strain localisation and inter- or intragranular failure are dependent on many factors that are ultimately microstructural in nature. The mechanisms are described and discussed in relation to reactor core materials. Full article
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14 pages, 976 KiB  
Review
Perspectives on a Severe Accident Consequences—10 Years after the Fukushima Accident
by Jinho Song
J. Nucl. Eng. 2021, 2(4), 398-411; https://doi.org/10.3390/jne2040030 - 11 Oct 2021
Cited by 4 | Viewed by 4061
Abstract
Scientific issues that draw international attention from the public and experts during the last 10 years after the Fukushima accident are discussed. An assessment of current severe accident analysis methodology, impact on the views of nuclear reactor safety, dispute on the safety of [...] Read more.
Scientific issues that draw international attention from the public and experts during the last 10 years after the Fukushima accident are discussed. An assessment of current severe accident analysis methodology, impact on the views of nuclear reactor safety, dispute on the safety of fishery products, discharge of radioactive water to the ocean, status of decommissioning, and needs for long-term monitoring of the environment are discussed. Full article
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Other

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10 pages, 2037 KiB  
Opinion
Nature, Energy and Society—A Scientific Study of the Options Facing Civilisation Today
by Wade Allison
J. Nucl. Eng. 2022, 3(3), 233-242; https://doi.org/10.3390/jne3030013 - 19 Aug 2022
Cited by 1 | Viewed by 3412
Abstract
The nations of the world plan to stop burning carbon fuels but have not fixed on any replacement. For social and economic confidence, they need to share a proper picture of the options. The science is simply explained and not in doubt, though [...] Read more.
The nations of the world plan to stop burning carbon fuels but have not fixed on any replacement. For social and economic confidence, they need to share a proper picture of the options. The science is simply explained and not in doubt, though widely misunderstood. Energy sources belong to three distinct groups: renewable, chemical and nuclear. Since human life began, it has adopted each of these in turn. In the past, initial disruptions have been more than off-set by the rise in human values that followed. Now, to complete the final step, we confront those who would look backward to the age of renewables. Instead, the world should look forward to a heavy dependence on nuclear energy with a confidence informed by natural science and openly shared in society. Full article
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