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15 pages, 449 KiB

Open AccessArticle

Optimal Axial-Probe Design for Foucault-Current Tomography: A Global Optimization Approach Based on Linear Sampling Method

by Brahim Benaissa, Samir Khatir, Mohamed Soufiane Jouini and Mohamed Kamel Riahi

Energies 2023, 16(5), 2448; https://doi.org/10.3390/en16052448 - 3 Mar 2023

Cited by 12 | Viewed by 2123

Abstract

This paper is concerned with the optimal design of axial probes, commonly used in the Non-Destructive Testing (NDT) of tube boiling in steam generators. The goal is to improve the low-frequency Foucault-current imaging of these deposits by designing a novel probe. The approach [...] Read more.

This paper is concerned with the optimal design of axial probes, commonly used in the Non-Destructive Testing (NDT) of tube boiling in steam generators. The goal is to improve the low-frequency Foucault-current imaging of these deposits by designing a novel probe. The approach uses a combination of an inverse problem solver with global optimization to find the optimal probe characteristics by minimizing a function of merit defined using image processing techniques. The evaluation of the function of merit is computationally intensive and a surrogate optimization approach is used, incorporating a multi-particle search algorithm. The proposed design is validated through numerical experiments and aims to improve the accuracy and efficiency of identifying deposits in steam generator tubes. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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27 pages, 11382 KiB

Open AccessArticle

Comparative Performance Evaluation of Gas Brayton Cycle for Micro–Nuclear Reactors

by Sungwook Choi, In Woo Son and Jeong Ik Lee

Energies 2023, 16(4), 2065; https://doi.org/10.3390/en16042065 - 20 Feb 2023

Cited by 2 | Viewed by 2606

Abstract

Gas Brayton cycles have been considered the next promising power cycles for microreactors. Especially the open-air and closed supercritical CO₂ (S-CO₂) Brayton cycles have received attention due to their high thermal efficiency and compact component sizes when compared to the [...] Read more.

Gas Brayton cycles have been considered the next promising power cycles for microreactors. Especially the open-air and closed supercritical CO₂ (S-CO₂) Brayton cycles have received attention due to their high thermal efficiency and compact component sizes when compared to the steam Rankine cycle. In this research, the performances of the open-air and closed S-CO₂ Brayton cycle at microreactor power range are compared with polytropic turbomachinery efficiency. When optimizing the cycle, three different optimization parameters are considered in this paper: maximum efficiency, maximum cycle specific work, and maximum of the product of both indicators. For the air Brayton cycle, the maximum of the product of both indicators allows to consider both efficiency and specific work while optimizing the cycle. However, for the S-CO₂ Brayton cycle, the best performing conditions follow either maximum efficiency or the maximum cycle specific work conditions. In general, the S-CO₂ power cycle should be designed and optimized to maximize the cycle specific work for commercial-scale application. The results show that the air Brayton cycle can achieve near 45% efficiency when it can couple with a microreactor with a core outlet temperature higher than 700 °C. However, the S-CO₂ power cycle can still achieve above 30% efficiency when it is coupled with a microreactor with a core outlet temperature higher than 500 °C, whereas the air Brayton cycle cannot even reach breakeven condition. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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16 pages, 5877 KiB

Open AccessArticle

Chromium-Coated Zirconium Cladding Neutronics Impact for APR-1400 Reactor Core

by Mohammad Alrwashdeh and Saeed A. Alameri

Energies 2022, 15(21), 8008; https://doi.org/10.3390/en15218008 - 28 Oct 2022

Cited by 37 | Viewed by 3055 | Correction

Abstract

The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium [...] Read more.

The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium fuel system (Zr–U) is currently used as a standard fuel system in pressurized water reactors around the world. This investigation presents the benefits of utilizing an alternative cladding material such as chromium coating and the effects on the thermal neutron parameters of the way in which the chromium coating is introduced in the reactor fuel. Among these significant benefits is an increase in the reactor fuel’s thermal conductivity, which improves reactor safety. Two types of fuel-cladding systems were investigated: Zircaloy–uranium (Zr–U) and Zircaloy–chromium (Zr–Cr–U) coating fuel systems. Neutronics analysis evaluations were performed for the selected fuel assemblies and a two-dimensional full core based on an APR-1400 reactor design. Neutronics analyses were performed for the application of the new fuel-cladding material systems using the reactor-physics Monte Carlo code Serpent 2.31. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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14 pages, 1622 KiB

Open AccessArticle

Analysis of a Hybrid Nuclear Renewable Energy Resource in a Distributed Energy System for a Rural Area in Nigeria

by Ronke M. Ayo-Imoru, Ahmed A. Ali and Pitshou N. Bokoro

Energies 2022, 15(20), 7496; https://doi.org/10.3390/en15207496 - 12 Oct 2022

Cited by 9 | Viewed by 2036

Abstract

Climate change is one of the global issues being combatted in recent times. One of the measures is a worldwide cutdown on carbon emissions. This has brought about the rapid development of technologies that can best actualise this goal. The decentralised energy system [...] Read more.

Climate change is one of the global issues being combatted in recent times. One of the measures is a worldwide cutdown on carbon emissions. This has brought about the rapid development of technologies that can best actualise this goal. The decentralised energy system is designed to harness the strengths of small power-generating sources such as renewable energy sources in a noncentralised manner to help meet the global need for clean energy. Renewable energy sources are faced with the challenge of intermittency, which brings about instability in the grid. Another source of clean energy is nuclear energy, which is traditionally large and not flexible; however, the recent development of technology has resulted in a scaled-down version of the large nuclear plants that are more flexible yet provide clean and stable electricity. This paper explores the possibility of deploying nuclear microreactors in the decentralised energy system and describes the features and the challenges of a decentralised energy system. The features of the small modular reactor that make it a viable candidate for the generating source in the decentralised energy system are explored. A case study for a DES system with a microreactor was conducted for a rural area in Nigeria. The HOMER software was used in simulating the optimum system, while TOPSIS was used in ranking the systems. The result showed that the PV/nuclear/battery system ranked first, followed by the PV/nuclear/wind and battery system. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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14 pages, 4177 KiB

Open AccessArticle

Effect of Temperature Gradient on Chloride Ion Diffusion in Nuclear Reactor Containment Building Concrete

by Boohyun An, Pyungyeon Cho, Remilekun A. Shittu, Tae-Yeon Kim, Paul Rostron, Akram AlFantazi and Yongsun Yi

Energies 2022, 15(15), 5581; https://doi.org/10.3390/en15155581 - 1 Aug 2022

Cited by 3 | Viewed by 1535

Abstract

Chloride diffusion through concrete is influenced by harsh environmental conditions such as high ambient temperature and relative humidity. This paper examined the influence of temperature gradient on chloride diffusion in concrete under high ambient temperature conditions. Chloride diffusion tests using cylindrical concrete samples [...] Read more.

Chloride diffusion through concrete is influenced by harsh environmental conditions such as high ambient temperature and relative humidity. This paper examined the influence of temperature gradient on chloride diffusion in concrete under high ambient temperature conditions. Chloride diffusion tests using cylindrical concrete samples were performed in constant temperature and temperature gradient conditions. In a temperature gradient condition, a much higher chloride concentration was measured than at constant temperatures, which could not be explained only by the mass diffusion driven by the concentration gradient. A new analytical model of chloride diffusion with the mass diffusion term including the temperature effect and the thermo-diffusion term including the temperature gradient effect was applied to the results, which showed that the thermo-diffusion contribution was significant. Using the analytical model with the mass diffusion (D_Cl) and thermo-diffusion (D_T) coefficients, the service life of reactor containment buildings (RCBs) in nuclear power plants (NPPs) in the Middle Eastern and North African (MENA) region was estimated. The results showed that the service life of the RCBs could be reduced by the temperature gradient, indicating the possible application of the proposed analytical model. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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19 pages, 6749 KiB

Open AccessArticle

Sensitivity Analysis of Ex-Vessel Corium Coolability Models in MAAP5 Code for the Prediction of Molten Corium–Concrete Interaction after a Severe Accident Scenario

by Muritala Alade Amidu, Yacine Addad and Akihide Hidaka

Energies 2022, 15(15), 5370; https://doi.org/10.3390/en15155370 - 25 Jul 2022

Cited by 3 | Viewed by 2929

Abstract

A postulated progressing severe accident scenario has been simulated using MAAP5 code with the focus on ex-vessel cooling of molten corium in the reactor cavity. Various parameters associated with the prediction of molten corium–concrete interaction (MCCI) are identified. Accordingly, a sensitivity analysis is [...] Read more.

A postulated progressing severe accident scenario has been simulated using MAAP5 code with the focus on ex-vessel cooling of molten corium in the reactor cavity. Various parameters associated with the prediction of molten corium–concrete interaction (MCCI) are identified. Accordingly, a sensitivity analysis is performed to assess the impact of these parameters on the predicted cavity floor erosion depth during this MCCI postulated accident. The sensitivity index of each variable parameter is determined using the Cotter indices method and Sobol′ indices method. At the early stage of the accident, the predicted cavity floor erosion depth is found to be highly sensitive to the downward heat transfer coefficient parameter with Cotter and Sobol′ indices of 94% and 50%, respectively. At the late phase of the accident, however, the cavity floor erosion depth becomes sensitive to melt eruption (Cotter index of 40%), water ingression (Cotter index of 13%), and particulate bed (Cotter index of 15%) parameters alongside the downward heat transfer coefficient (Cotter index of 16%) with the melt eruption parameter becoming dominant. Thus, the sensitivity of the code′s predictions can be minimized by improving the physical models associated with these parameters. Moreover, the sensitivity indices of these parameters can be used by model developers to identify unimportant parameters in a bid to reduce the dimension of the problem with the aim of improving the current predictive capabilities to conduct MCCI-related safety analyses. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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20 pages, 15206 KiB

Open AccessArticle

Neutronic Analysis of SiC/SiC Sandwich Cladding Design in APR-1400 under Normal Operation Conditions

by Maithah M. Alaleeli, Saeed A. Alameri and Mohammad Alrwashdeh

Energies 2022, 15(14), 5204; https://doi.org/10.3390/en15145204 - 18 Jul 2022

Cited by 15 | Viewed by 2974

Abstract

Our aim is to study the neutronic behaviour of potential accident-tolerant fuel (ATF) claddings in a pressurised water reactor under normal operations. This work compares ATF silicon carbide composite (SiC/SiC) cladding to conventional ZIRLO^TM cladding in APR-1400. Additionally, a “sandwich” cladding design [...] Read more.

Our aim is to study the neutronic behaviour of potential accident-tolerant fuel (ATF) claddings in a pressurised water reactor under normal operations. This work compares ATF silicon carbide composite (SiC/SiC) cladding to conventional ZIRLO^TM cladding in APR-1400. Additionally, a “sandwich” cladding design developed by the CEA is used for SiC/SiC. The design structure includes a liner in between two layers of the composite to ensure leak tightness. The two proposed liners are Niobium (Nb) and Tantalum (Ta). Serpent 2, a Monte Carlo reactor physics lattice code, is employed to model both cladding materials in APR-1400 at three different levels: pin cell, fuel assembly, and core. The criticality, neutron spectrum, actinide inventory, and power distribution as a function of burnup are investigated. The simulations show that SiC/SiC with the Nb liner displays a far superior performance than the Ta liner across all examined characteristics. Ta leads to a harder neutron spectrum and increased Pu-239 content throughout the cycle, while Nb presents negligible effects. In fact, SiC/SiC with the Nb liner performs very similarly to ZIRLO^TM at all model levels. The results indicate that, in terms of neutronics, the adoption of the SiC/SiC composite would entail little to no changes to current APR-1400 operations. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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Review

Jump to: Research

24 pages, 1941 KiB

Open AccessReview

Degradation of Concrete Structures in Nuclear Power Plants: A Review of the Major Causes and Possible Preventive Measures

by Pathath Abdul Rasheed, Sunitha K. Nayar, Imad Barsoum and Akram Alfantazi

Energies 2022, 15(21), 8011; https://doi.org/10.3390/en15218011 - 28 Oct 2022

Cited by 16 | Viewed by 4918

Abstract

Concrete, an integral part of a nuclear power plant (NPP), experiences degradation during their operational lifetime of the plant. In this review, the major causes of concrete degradation are extensively discussed including mechanisms that are specific to NPPs. The damage mechanism could be [...] Read more.

Concrete, an integral part of a nuclear power plant (NPP), experiences degradation during their operational lifetime of the plant. In this review, the major causes of concrete degradation are extensively discussed including mechanisms that are specific to NPPs. The damage mechanism could be chemical or physical. The major causes of chemical degradation include alkali–aggregate reactions, leaching, sulfate attack, bases and acids attack, and carbonation. Physical degradation is a consequence of both environmental and mechanical factors combined. These factors are mainly elevated temperature, radiation, abrasion and erosion, salt crystallization, freeze–thaw distortions, fatigue and vibration. Additionally, steel reinforcements, prestressing steels, liner plates, and structural steel also experience degradation. The prospective areas in the structural components of the NPP where the degradation could occur are mentioned and the effective solutions to the causes of degradation are highlighted. These solutions are designed to enhance the physical and chemical characteristics of concrete. Some of the major recommendations include addition of mineral substitutes, use of low water-to-cement ratio as well as low water-to-binder ratio, use of low alkali cement, use of special aggregates and fibers, use of corrosion inhibitors, use of cathodic protection, etc. The review concludes with an overview of present methods and possible recommendations used to enhance the quality of concrete towards preventing concrete degradation and increasing the lifetime of NPPs. Full article

(This article belongs to the Special Issue New Challenges in Nuclear Energy Systems)

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