Nuclear Engineering and Design最新文献_第2页

Comparative neutronic analysis of diverse seed blanket unit (SBU) concepts for the ACP-100 small modular reactor ACP-100小型模块化反应堆不同种籽毯单元（SBU）概念的比较中子分析

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114507

H. Rainad Khan Rohan , Md. Abidur Rahman Ishraq , Anton Evgenievich Kruglikov

{"title":"Comparative neutronic analysis of diverse seed blanket unit (SBU) concepts for the ACP-100 small modular reactor","authors":"H. Rainad Khan Rohan , Md. Abidur Rahman Ishraq , Anton Evgenievich Kruglikov","doi":"10.1016/j.nucengdes.2025.114507","DOIUrl":"10.1016/j.nucengdes.2025.114507","url":null,"abstract":"<div><div>This study investigates the neutronic feasibility of diverse seed blanket unit (SBU) concepts for the ACP-100 small modular reactor using the Monte Carlo code SERPENT. Pairing four different seed fuels (UO<sub>2</sub>, UC, UN, and MOX) with three blanket materials (natural UO<sub>2</sub>, depleted UO<sub>2</sub>, and mixed ThO<sub>2</sub>-UO<sub>2</sub>), a total of twelve different SBU-utilizing core models were developed and evaluated. The results indicate that MOX fuels with depleted/natural UO<sub>2</sub> blankets are unsuitable for SBU cores as they cannot sustain prolonged core operation. Of the twelve models, eight successfully remained critical for more than one effective full power year (1 EFPY). The best performing seed/blanket compositions were UC/Th and UO<sub>2</sub>/Th with cycle lengths of 1.88 EFPY and 1.75 EFPY, and discharge burnups of 15.85 MWd/kg and 15.39 MWd/kg respectively. The discharge burnup and cycle length of UC seed paired with thorium blanket surpassed those of the reference by 7.9% and 1.6% respectively. All SBU cores achieved high conversion ratios, maintained negative fuel and moderator temperature coefficients (FTC and MTC) and had lower power peaking factors compared to the reference, indicating a more uniform power distribution. The SBU cores utilizing thorium had more negative temperature coefficients than the reference and produced less fissile plutonium (with the exception of the UN-seeded core) at the end of one year. Although a degradation in beta effective was observed, all SBU cores had sufficient control rod reactivity worth to render them subcritical at any instant during operation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114507"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266803","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Large eddy simulation on the flow and thermal fluctuation of non-isothermal water mixing in reactor upper plenum 反应器上充气室非等温水混合流动及热波动的大涡模拟

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114514

Linghao Liu , Guo-Yan Zhou , Xueyao Xiong , Xing Luo , Shan-Tung Tu

{"title":"Large eddy simulation on the flow and thermal fluctuation of non-isothermal water mixing in reactor upper plenum","authors":"Linghao Liu , Guo-Yan Zhou , Xueyao Xiong , Xing Luo , Shan-Tung Tu","doi":"10.1016/j.nucengdes.2025.114514","DOIUrl":"10.1016/j.nucengdes.2025.114514","url":null,"abstract":"<div><div>In nuclear reactors, temperature fluctuations resulting from the mixing of non-isothermal fluids in the upper plenum may induce fatigue damage to adjacent structures, a phenomenon known as thermal striping. To accurately capture temperature fluctuations in the coolant and surrounding components, we employed Large Eddy Simulation (LES) and developed a simplified hexagonal-jet model for simulating these fluctuations within the actual reactor core structure. The results were compared with those from the existing coaxial-jet model. The results show that the temperature fluctuation of the fluid under the influence of the hexagonal-jet progressively increases in the axial direction, while it significantly attenuates at the lower surface of the plate. This temperature fluctuation arises from the mixing of the jet with the crossflow. In contrast, the temperature fluctuation induced by the coaxial-jet results from the intrusion of hot fluid into the cold fluid and their subsequent mixing. The intensity of temperature fluctuations for the two models varies differently in the radial and axial directions. Furthermore, the temperature signal of the fluid influenced by the coaxial-jet is more concentrated, with a larger amplitude compared to that of the hexagonal-jet. Consequently, the results obtained by the coaxial-jet model can be regarded as a conservative estimation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114514"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266806","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Dissolution experiments with surrogate treated organic wastes immobilized by hot isostatic pressing 热等静压固定化有机废物的溶解实验

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114533

Karine Ferrand, Pieter Schroeders, Sébastien Caes, Karel Lemmens

{"title":"Dissolution experiments with surrogate treated organic wastes immobilized by hot isostatic pressing","authors":"Karine Ferrand, Pieter Schroeders, Sébastien Caes, Karel Lemmens","doi":"10.1016/j.nucengdes.2025.114533","DOIUrl":"10.1016/j.nucengdes.2025.114533","url":null,"abstract":"<div><div>The presence of organic compounds in radioactive wastes might be incompatible with long-term waste management options. Consequently, as part of the H2020-PREDIS project, different thermal treatment routes and conditioning techniques were first used to process surrogates of solid organic wastes. Then, the chemical durability of the conditioned wastes was investigated performing dissolution experiments in a synthetic cementitious water at pH12.7, 22 °C and in diluted conditions. These tests were conducted with a glass–ceramic, produced by hot isostatic pressing using a mixture of ashes and sodium tetraborate (95/5 wt%). The results showed that glass–ceramic dissolution was incongruent and controlled by diffusion. SEM–EDX analysis showed the presence of an altered zone of about 45 µm after 730 days of alteration, with amorphous and crystalline phases (spinel and chlorapatite), and containing voids due to the dissolution of the glass matrix. The average diffusion coefficients were in the range of (4.26 ± 1.36) × 10<sup>−15</sup>  m<sup>2</sup>/s, with the lowest and highest values for Zn and B, respectively. This corresponds to leachability indexes in the range of 10.40 ± 0.14 and 8.54 ± 0.07. The maximum dissolution rate of the glass matrix based on the B release between the start of the tests and 28 days was similar to that found when borosilicate glass was corroded by a KOH solution at pH12.5. Hot isostatic pressing was thus successfully used to immobilize ashes, and the resulting glass–ceramic exhibited good chemical durability under alkaline conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114533"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266802","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Progressive Multiphysics analysis to Implement a ceramic gas gap thermal barrier Enabling High-Temperature irradiation of High-Burnup fuel rods 实现高燃耗燃料棒高温辐照的陶瓷气隙热障的渐进多物理场分析

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114536

Hai Wang , Qinglong Wen , Zhengang Duan , Zhongkai Mei

{"title":"Progressive Multiphysics analysis to Implement a ceramic gas gap thermal barrier Enabling High-Temperature irradiation of High-Burnup fuel rods","authors":"Hai Wang , Qinglong Wen , Zhengang Duan , Zhongkai Mei","doi":"10.1016/j.nucengdes.2025.114536","DOIUrl":"10.1016/j.nucengdes.2025.114536","url":null,"abstract":"<div><div>To address the irradiation testing requirements of high-temperature high-burnup fuel rods (HT-HBFR) in low-temperature, low-pressure research reactors, this study proposes an irradiation assembly incorporating a ceramic gas-gap thermal barrier (CGGTB) to mitigate localized overheating in the reactor core. Heat transfer simulations demonstrated that, within a heat flux range of 1,000 ± 200 W/m<sup>2</sup>, the maximum deviation between experimental and theoretical values was –3.76 %, confirming the engineering feasibility of the thermal barrier and it has high predictive accuracy under anticipated operational conditions. Further theoretical calculations, numerical simulations, and hydraulic experiments were conducted to optimize the throttling device geometry and coolant channel parameters. At a target flow rate of 0.854 kg/s, the theoretical pressure drop of the downward coolant flow through the irradiation device exceeded the experimental value by 9.46 %. Progressive multiphysics simulations systematically validated the decoupling control capability of the CGGTB over fuel pellet temperature, cladding surface temperature, and coolant interface temperature. Under a rated heat flux of 1,000 kW/m<sup>2</sup>, the maximum fuel pellet temperature reached 1,259 ℃ and the cladding surface temperature was 615 ℃, both satisfying the test requirements, while the coolant outlet temperature remained at 53.3 ℃, well below the thermal safety limit (195 ℃) of the High Flux Engineering Test Reactor (HFETR). This irradiation strategy provides high-precision technical support for the performance verification of HT-HBFR in non-prototype reactor environments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114536"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266805","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Nuclear waste for hydrogen production: methods, advantages, and future perspectives 核废料制氢：方法、优势和未来展望

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-09 DOI: 10.1016/j.nucengdes.2025.114511

Shatha Alyazouri , Muhammad Zubair

{"title":"Nuclear waste for hydrogen production: methods, advantages, and future perspectives","authors":"Shatha Alyazouri , Muhammad Zubair","doi":"10.1016/j.nucengdes.2025.114511","DOIUrl":"10.1016/j.nucengdes.2025.114511","url":null,"abstract":"<div><div>Hydrogen has become a promising energy carrier as the need for sustainable and clean energy sources increases globally. Utilizing nuclear waste is a novel method of producing hydrogen that transforms a persistent environmental issue into a useful resource. The present work provides a comprehensive review of innovative approaches for hydrogen production through the effective utilization of nuclear waste. Based on the existing research, it was found that nuclear waste can significantly enhance hydrogen generation through a variety of advanced methods, including catalyst-enhanced electrolysis, methane reforming, and thermochemical cycles. Other promising techniques involve radiation-enhanced electrolysis cells, feeding radioactive waste into a heater to generate electricity for powering electrolysis cells, radiolysis, and liquid plasma photocatalysis. These techniques have several advantages, including lowering the amount of radioactive waste, lowering the requirement for long-term storage, and supplying a steady supply of hydrogen. Moreover, the limitations and challenges associated with these methods have been thoroughly explored, including the risk of syngas contamination, chemical modification of the catalyst, and stringent regulations that hinder research progress in this field.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114511"},"PeriodicalIF":2.1,"publicationDate":"2025-10-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266809","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

PARCS pin-wise simulation with a cross section correction system based on the Super-Homogenisation method 基于超均匀化方法的截面校正系统的PARCS引脚方向仿真

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-08 DOI: 10.1016/j.nucengdes.2025.114515

Kanglong Zhang, Luigi Mercatali, Victor Hugo Sanchez-Espinoza

{"title":"PARCS pin-wise simulation with a cross section correction system based on the Super-Homogenisation method","authors":"Kanglong Zhang, Luigi Mercatali, Victor Hugo Sanchez-Espinoza","doi":"10.1016/j.nucengdes.2025.114515","DOIUrl":"10.1016/j.nucengdes.2025.114515","url":null,"abstract":"<div><div>A cross-section (XS) correction system was developed at the Karlsruhe Institute of Technology (KIT) for pin-wise simulations in PARCS, utilizing the SuPer-Homogenization (SPH) method in an iterative Python-based framework. The system uses Monte Carlo Serpent2 pin-wise solutions as a reference to correct homogenized pin-wise XS, improving PARCS’ accuracy in predicting neutron reaction rates and pin-wise power distributions. Verification was conducted with four test cases of increasing complexity: (a) a 3x3 mini assembly, (b) KONVOI reactor fuel assemblies, (c) a 3x3 mini-core, and (d) the Karlsruhe Small Modular Reactor (KSMR). Results show that the corrected XS significantly enhances PARCS’ diffusion solver accuracy, closely matching Serpent2 results, in pin-by-pin core simulations. Notably, this approach outperforms traditional assembly-wise Pin Power Reconstruction (PPR), particularly in central core regions. The key outcome of this work is that, for PARCS pin-wise simulations, computational accuracy is enhanced, making the prediction of local safety parameters feasible, especially if coupled with a Thermal-Hydraulic (TH) code, such as a sub-channel code. Furthermore, this work extends PARCS’s capability to perform pin-wise simulations with the standard nodal solver.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114515"},"PeriodicalIF":2.1,"publicationDate":"2025-10-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266825","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Advanced treatment strategies for challenging radioactive Wastes: Recent Developments and future directions 具有挑战性的放射性废物的先进处理策略：最新发展和未来方向

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-08 DOI: 10.1016/j.nucengdes.2025.114501

Elena Torres Álvarez , Michael Gracey , Mark Cowper , Quoc Tri Phung , Borys Zlobenko , Federica Pancotti , Mafalda Guerra , Märten Kala , Janno Torop , Alan H. Tkaczyk

{"title":"Advanced treatment strategies for challenging radioactive Wastes: Recent Developments and future directions","authors":"Elena Torres Álvarez , Michael Gracey , Mark Cowper , Quoc Tri Phung , Borys Zlobenko , Federica Pancotti , Mafalda Guerra , Märten Kala , Janno Torop , Alan H. Tkaczyk","doi":"10.1016/j.nucengdes.2025.114501","DOIUrl":"10.1016/j.nucengdes.2025.114501","url":null,"abstract":"<div><div>The management of low- and intermediate-level radioactive waste (LILW) presents increasing challenges, particularly for waste streams with complex chemistries, variable radioactivity, or combined toxic and radiological risks. Conventional treatment methods, while effective for conventional waste types, are often inadequate for these problematic streams, which require advanced and tailored approaches. This review evaluates recent progress in thermal (pyrolysis, plasma processes, fluidized bed steam reforming, molten salt oxidation) and chemical technologies (wet and electrochemical oxidation) for treating such wastes. The novelty of this work lies in its integrated analysis, which combines comparative technology mapping with assessments of scalability, regulatory and societal acceptance, and the integration of lifecycle environmental and economic metrics. Cross-technology performance matrices are presented to support process selection, together with a strategic research roadmap that prioritizes secondary waste minimization, flexible acceptance criteria, and the inclusion of Life Cycle Assessment (LCA). The study also highlights future trends such as mobile modular treatment systems, hybrid technologies that integrate multiple treatment approaches, and the increasing use of artificial intelligence (AI) and Internet of Things (IoT) technologies for real-time monitoring and process optimization. Overall, this review aims to provide a practical framework to guide researchers, industry, and regulators toward safe, sustainable, and industrially viable solutions for complex radioactive waste management.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114501"},"PeriodicalIF":2.1,"publicationDate":"2025-10-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267568","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Thermal-hydraulic analysis of temporary spent fuel storage in the upper pool of the Kuosheng units under postulated leakage accident 假定泄漏事故下阔胜机组上池临时乏燃料库热水力分析

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-07 DOI: 10.1016/j.nucengdes.2025.114527

Yen-Shu Chen

{"title":"Thermal-hydraulic analysis of temporary spent fuel storage in the upper pool of the Kuosheng units under postulated leakage accident","authors":"Yen-Shu Chen","doi":"10.1016/j.nucengdes.2025.114527","DOIUrl":"10.1016/j.nucengdes.2025.114527","url":null,"abstract":"<div><div>Two BWR/6 units at the Kuosheng Nuclear Power Plant are undergoing decommissioning. Since 2018, the cask loading pool (CLP), which is a sub-pool of the spent fuel pool (SFP), has been used for spent fuel storage. To prepare for the implementation of a dry-cask storage facility, the plant has committed to restore the CLP to its original state. During the recovery period, the assemblies stored in the CLP must be temporarily moved to the upper pool (UP), which is a part of the Mark III containment. In this study, a thermal–hydraulic model has been developed using the GOTHIC code for the analysis of postulated leakage accidents. Effects of the leakage size, UP spray mitigation, reactor building (RB) modeling, and higher decay power are investigated. While the large leakage results in faster water inventory loss, it also promotes earlier air circulation in the fuel region to suppress the fuel temperature rise. The Kuosheng plant has installed spray mitigation above the UP, and the calculation results show that the spray effect can sufficiently prevent the fuel from being overheated. Three-dimensional RB modeling demonstrates good air mixing within this space. Additionally, cases with higher decay power are calculated. The availability of spray mitigation ensures the maintenance of fuel integrity.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114527"},"PeriodicalIF":2.1,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266823","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Study on the thermal hydraulic performance of a separated structure shell-and-tube steam generator under various operating conditions 分离式管壳式蒸汽发生器在不同工况下的热工性能研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-07 DOI: 10.1016/j.nucengdes.2025.114505

Huaishuang Shao, Jiang Chang, Jian Jiao, Zhiyuan Liang, Yungang Wang, Qinxin Zhao

{"title":"Study on the thermal hydraulic performance of a separated structure shell-and-tube steam generator under various operating conditions","authors":"Huaishuang Shao, Jiang Chang, Jian Jiao, Zhiyuan Liang, Yungang Wang, Qinxin Zhao","doi":"10.1016/j.nucengdes.2025.114505","DOIUrl":"10.1016/j.nucengdes.2025.114505","url":null,"abstract":"<div><div>A thermal–hydraulic performance analysis program based on Fluent was developed for separated structure shell-and-tube steam generator. The coupled heat transfer on both sides of the steam generator and the anisotropic flow resistance model of the vapor–liquid two-phase flow across tube bundle were mainly considered. The reliability of the program has been verified by experimental data. Based on the program, the effects of different inlet flue gas temperatures, flue gas mass flow rates and operating pressures on the thermal–hydraulic characteristics were studied. It was found that flue gas temperature and flue gas flow rate mainly affect the heat flux distribution. With the increase of flue gas temperature and flow rate, the vapor void fraction and flow velocity on the shell side increase accordingly. The flow resistance inside the shell and the local resistance at the inlet of the riser increase accordingly, but the overall natural circulation flow rate still increased by 38.72% and 13.19% respectively. It showed a positive compensation characteristic with head load. The operating pressure mainly affects the physical property of the working fluid. The lower the pressure, the higher the vapor void fraction on the shell side. Although the flow resistance inside the shell increased under low pressure condition, the overall natural circulation flow rate still increased by 42.33% due to larger circulation pressure difference. This can provide a reference for the structural design of the separated structure shell-and-tube steam generator.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114505"},"PeriodicalIF":2.1,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267623","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Comparison of thermal and structural performance of conventional concrete and UHPC for spent nuclear fuel storage structures 核废料贮存结构用常规混凝土和超高性能混凝土的热性能和结构性能比较

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-07 DOI: 10.1016/j.nucengdes.2025.114519

Nataliia Igrashkina , Mohamed A. Moustafa , Mustafa Hadj-Nacer

{"title":"Comparison of thermal and structural performance of conventional concrete and UHPC for spent nuclear fuel storage structures","authors":"Nataliia Igrashkina , Mohamed A. Moustafa , Mustafa Hadj-Nacer","doi":"10.1016/j.nucengdes.2025.114519","DOIUrl":"10.1016/j.nucengdes.2025.114519","url":null,"abstract":"<div><div>Spent nuclear fuel (SNF) is currently stored in a growing number of dry cask storage structures across the US. With concrete aging risks, licensing renewals challenges, and the absence of a permanent SNF repository, there is a pressing need for extending the service life of the existing storage systems and rethinking the design of new systems for longevity using durable materials. Ultra-high performance concrete (UHPC) possesses superior mechanical and durability properties and presents a promising solution for retrofitting existing or building new SNF storage facilities. Only very limited research is available on the possible utilization of UHPC in dry storage systems. As such, this study takes a first look at rethinking the design of current SNF reinforced concrete horizontal modules storage facilities using UHPC, and focuses on the comparative thermal and structural performance of conventional concrete and UHPC in a typical horizontal storage module. The study considers computational fluid dynamics (CFD) analysis to establish temperature distributions that are used subsequently in structural finite element analysis. Demands from various load combinations are compared to respective concrete and UHPC section capacities for assessment. The results show that UHPC components have a bigger reserved structural capacity and better thermal performance than conventional concrete, and could provide a significantly enhanced structural performance besides the outstanding durability and extended service life.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114519"},"PeriodicalIF":2.1,"publicationDate":"2025-10-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267625","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0