Annals of Nuclear Energy最新文献

Numerical modeling and validation for transport behaviors of tracer gas in the European tracer experiment (ETEX-1) 欧洲示踪实验（ETEX-1）中示踪气体输运行为的数值模拟与验证

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111953

Zhenhui Ma, Xiuhuan Tang, Longbo Liu, Lihong Bao, Baosheng Wang, Tengyue Ma, Pan Hu, Yonggang Zhangsun, Chunlei Su

{"title":"Numerical modeling and validation for transport behaviors of tracer gas in the European tracer experiment (ETEX-1)","authors":"Zhenhui Ma, Xiuhuan Tang, Longbo Liu, Lihong Bao, Baosheng Wang, Tengyue Ma, Pan Hu, Yonggang Zhangsun, Chunlei Su","doi":"10.1016/j.anucene.2025.111953","DOIUrl":"10.1016/j.anucene.2025.111953","url":null,"abstract":"<div><div>The European Tracer Experiment (ETEX-1) was conducted across Europe by European Commission, World Meteorological Organization and International Atomic Energy Agency jointly on October 26th, 1994. In the experiment, PMCH (Perflouro-Methyl-Cyclo-Hexane) was adopted as tracer gas, and it was released from Monterfil, France for nearly 24 hours. Totally 168 observation stations for PMCH were settled in different countries in Europe to observe time-varying concentration for arriving tracer gas. This study established a numerical method for simulating tracer gas transport using the meteorology-air quality coupling model WRF-CMAQ and applied this method to simulate tracer gas transport during ETEX-1. On the basis of simulation results for PMCH concentration in atmosphere, transport characteristics can be analyzed in detail. According to simulation results, 72 hours after release, tracer gas could influence most part of central and southeast Europe. Furthermore, gas concentrations at different times from simulation results were compared with data observed from representative observation stations. Finally, a comprehensive statistical analysis was performed to evaluate the model’s performance in meso-scale atmospheric transport simulation. According to analysis results for statistical indicators, although the value of FB was slightly lower than the recommended distribution range (FB = −0.308, recommended distribution range: −0.3 ∼ 0.3), Correlational Coefficient (r = 0.275, recommended distribution range: −1.0 ∼ 1.0) and NMSE (NMSE = 1.26, recommended distribution range: <4.0) were both in the acceptable distribution range, which means the present model is comparable with other models, even show better performance in analysis for some statistical indicators.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111953"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322536","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}

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Fuel performance modeling of used nuclear fuel transfer between storage pools and its impact on dry storage 乏燃料在贮存池间转移的燃料性能建模及其对干贮存的影响

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111934

Piotr Konarski, Cedric Cozzo, Grigori Khvostov, Hakim Ferroukhi

{"title":"Fuel performance modeling of used nuclear fuel transfer between storage pools and its impact on dry storage","authors":"Piotr Konarski, Cedric Cozzo, Grigori Khvostov, Hakim Ferroukhi","doi":"10.1016/j.anucene.2025.111934","DOIUrl":"10.1016/j.anucene.2025.111934","url":null,"abstract":"<div><div>The transfer of fuel in dry casks between the storage pools at the Swiss nuclear power plant Gösgen is known as the Wet-To-Dry-To-Wet (W2D2W) transfer procedure. The process is associated with vacuum drying, transfer to an external storage pool and reflooding. The high cladding temperature and stress during the drying and transport in combination with quick reflooding can lead to precipitation of radial hydrides and deterioration of the cladding mechanical properties. In this report, the impact of the drying, transfer and reflooding periods on the concentration and distribution of radial hydrides is studied. The most limiting rods are identified based on the criteria proposed in this work and submitted to multiple scenarios considering various durations of the W2D2W stages. The results show the impact of different phases and reveal that the reflooding speed/rate has the biggest influence on the hydrogen behavior. The analysis of the simulated dry storage scenarios shows that the maximum temperature obtained at the end of drying governs the hydrogen migration into the liner.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111934"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322537","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}

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Optimisation of waste management from the remediation of uranium contaminated soils 铀污染土壤修复中废物管理的优化

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111955

A. Álvarez, J.A. Suárez-Navarro, M.A.Lopez Ponte, V.M. Expósito-Suárez

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Secondary side system design: A new calculation method for passive residual heat removal of liquid fuel molten salt reactors 二次侧系统设计：液体燃料熔盐堆被动余热排出的一种新的计算方法

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111946

Shuaiyu Xue , Chong Zhou , Pinyan Huang , Yang Zou

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Feasibility Assessment of Thorium/Plutonium-Based Accident-Tolerant fuel in an Innovative fuel assembly design for the VVER-1200 reactor 基于钍/钚的耐事故燃料在VVER-1200反应堆创新燃料组件设计中的可行性评估

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111954

Nassar Alnassar , Shlash A. Luhaib , Mohamed Y.M. Mohsen , Mostafa M.A. Khater , A. Abdelghafar Galahom

{"title":"Feasibility Assessment of Thorium/Plutonium-Based Accident-Tolerant fuel in an Innovative fuel assembly design for the VVER-1200 reactor","authors":"Nassar Alnassar , Shlash A. Luhaib , Mohamed Y.M. Mohsen , Mostafa M.A. Khater , A. Abdelghafar Galahom","doi":"10.1016/j.anucene.2025.111954","DOIUrl":"10.1016/j.anucene.2025.111954","url":null,"abstract":"<div><div>Radioactivity reduction and effective waste management remain critical challenges for the nuclear industry, affecting long-term safety and public acceptance. This study explores the feasibility of employing thorium–plutonium accident-tolerant fuels (ATFs) in a novel blanket–seed (BS) assembly for the VVER-1200 reactor. Comparative evaluations were conducted based on fuel burnup, thermal power distribution, key safety-related parameters, and radioactivity analysis. Also, the decay of spent fuel radioactivity during the cooling process was assessed for fuel stored in a pool containing borated water with a boron concentration of 600 pcm. Results indicate that all BS configurations achieve extended reactivity lifetimes (up to ∼ 1750 EFPDs), and enhanced neutron economy due to higher atomic densities in denser fuels. The BS assembly consumed a significant amount of the plutonium used in the seed rods. Compared to UO<sub>2</sub>, thorium-based fuels exhibit significantly reduced total actinide and non-actinide radioactivity, demonstrating a favorable trade-off for long-term waste management.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111954"},"PeriodicalIF":2.3,"publicationDate":"2025-10-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322535","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}

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Multi-attribute evaluation method of nuclear island plant overall layout schemes based on linguistic intuitionistic fuzzy VIKOR 基于语言直觉模糊VIKOR的核岛电厂总体布局方案多属性评价方法

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-18 DOI: 10.1016/j.anucene.2025.111950

Dong Hao , JinCheng Su, YanFang Fan

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Comparative analysis of CANDU severe accident using CAISER 应用CAISER对CANDU严重事故进行对比分析

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111931

Jun-young Kang Ph.D., Yong Mann Song, Dong Gun Son, Keun Sang Choi, Jong Yeob Jung, Jun Ho Bae, Byeonghee Lee

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Revisiting mini-max polynomial approximation method for nuclear fuel depletion calculation 再论核燃料耗竭计算的最小-最大多项式近似法

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111948

Go Chiba , Kento Yamamoto , Hiroaki Nagano

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Analysis of the effect of LFR bending fuel assembly on thermal–hydraulic characteristics LFR弯曲燃料组件对热工特性的影响分析

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111933

JiTong Sun , Mei Huang , YaoDi Li , AoNan Li , YanTing Cheng

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An adaptive three-dimensional model for thermal hydraulics analysis of a PWR steam generator 一种用于压水堆蒸汽发生器热水力分析的自适应三维模型

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-10-17 DOI: 10.1016/j.anucene.2025.111922

Yung Suk Nam , Han Young Yoon

{"title":"An adaptive three-dimensional model for thermal hydraulics analysis of a PWR steam generator","authors":"Yung Suk Nam , Han Young Yoon","doi":"10.1016/j.anucene.2025.111922","DOIUrl":"10.1016/j.anucene.2025.111922","url":null,"abstract":"<div><div>Pressurized Water Reactor (PWR) steam generators (SGs) feature complex internal structures, including numerous heat exchanger tubes and support components. Accurate modeling of these internal structures is essential for reliable thermal–hydraulic analysis. However, conventional modeling approaches for SG thermal hydraulic analysis often require significant effort and time, limiting efficiency and flexibility. In this paper, we propose an adaptive three-dimensional (3D) modeling approach for the internal structures of PWR SGs, enabling comprehensive and efficient thermal–hydraulic analysis applicable to various designs. The 3D mesh encompasses all critical regions of the SG, including the U-tube heat exchanger, riser, downcomer, and steam dome. A dedicated mesh model accurately representing the shape of the U-tube bundle was developed, with the total tube length calculated from the model differing by less than 1% compared to design specifications. The CUPID-SG code (CUPID for SG application), based on the two-fluid model, is employed to perform the thermal–hydraulic analysis using the proposed 3D mesh. CUPID-SG incorporates primary coolant, U-tube heat conduction, and secondary coolant two-phase flow models, providing a comprehensive framework for simulation. The effectiveness of the proposed approach is demonstrated through the thermal–hydraulic analysis of the APR1400 SG, validating its capability for accurate and adaptable modeling of PWR SGs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"227 ","pages":"Article 111922"},"PeriodicalIF":2.3,"publicationDate":"2025-10-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145322482","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