Nuclear Engineering and Technology最新文献_第8页

Analysis of double-ended guillotine break accident of surge line in ACP100 based on coupling method: Development and validation of the coupling method 基于耦合法分析 ACP100 中浪涌线的双端铡刀断裂事故：耦合方法的开发与验证

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-22 DOI: 10.1016/j.net.2024.07.046

Fan Miao, Bin Zhang, Tianci Xie, Hao Yang, Jianqiang Shan

{"title":"Analysis of double-ended guillotine break accident of surge line in ACP100 based on coupling method: Development and validation of the coupling method","authors":"Fan Miao, Bin Zhang, Tianci Xie, Hao Yang, Jianqiang Shan","doi":"10.1016/j.net.2024.07.046","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.046","url":null,"abstract":"The ACP100 is a small modular reactor (SMR) designed and built in China, featuring an integrated primary loop and passive safety systems. In SMRs, the interaction between the reactor coolant system, containment vessel, and other systems is highly interdependent. Therefore, it is necessary to use high-precision analysis codes, which can be achieved by coupling multiple codes. This paper is the first part of a study which investigates the LOCA in the ACP100 using a coupled platform. In this paper, a coupling platform was developed for transient analysis of SMR accidents, based on the system code NUSOL-SYS and the Integrated Severe Accident Analysis (ISAA) code. An inter-process communication module was developed adopting shared memory and event objects to exchange data between the two codes. A coupling interface defining the data to be exchanged was proposed. Both NUSOL-SYS and ISAA were modified to perform synchronous time step control. The coupling platform were validated through a hypothetical scenario and Edwards’ pipe blowdown experiment, demonstrating exact consistency and high accuracy. This coupling platform offers a new method for SMR accident analysis, providing a foundation for future work.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141784991","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

Preliminary neutronics design and analysis of lithium heat pipe cooled space reactor with low-enriched uranium 使用低浓缩铀的锂热管冷却空间反应堆的初步中子设计与分析

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-20 DOI: 10.1016/j.net.2024.07.042

Xiaoliang Zou, Yanting Sun, Qiusun Zeng, Xiaojian Wen, Xiaogang Cao, Xi Huang, Yibao Liu

{"title":"Preliminary neutronics design and analysis of lithium heat pipe cooled space reactor with low-enriched uranium","authors":"Xiaoliang Zou, Yanting Sun, Qiusun Zeng, Xiaojian Wen, Xiaogang Cao, Xi Huang, Yibao Liu","doi":"10.1016/j.net.2024.07.042","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.042","url":null,"abstract":"The space nuclear reactor cooled by heat pipes has become the preferred choice for future space missions and deep space exploration missions. The use of low-enriched uranium (LEU) is promoted to achieve the goal of nuclear non-proliferation worldwide. In this study, a lithium heat pipe cooled space reactor with LEU (HP-LEU) was proposed based on Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs), with the addition of moderators. The HP-LEU employs yttrium hydride (YH) as the moderator and 19.9 % enriched uranium nitride (UN) as the fuel. The neutronics analysis has been performed on the HP-LEU reactor and the results have showed that the HP-LEU has a lifetime of more than 12 years. Two control systems have been applied in the reactor and have demonstrated the capacity to independently regulate and shut down the reactor. The total temperature reactivity coefficients are consistently negative, indicating that the HP-LEU reactor is inherently safe during operation. During normal operation, the temperatures of the materials are all acceptable. This study can serve as a reference for lithium heat pipe cooled space reactors with LEU.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141784992","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

Lifetime thermal analysis of the CANDU spent fuel storage canister at the Wolsung site Wolsung 核电厂 CANDU 乏燃料贮存罐的寿命热分析

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-18 DOI: 10.1016/j.net.2024.07.041

Tae Gang Lee, Taehyung Na, Byongjo Yun, Jae Jun Jeong

{"title":"Lifetime thermal analysis of the CANDU spent fuel storage canister at the Wolsung site","authors":"Tae Gang Lee, Taehyung Na, Byongjo Yun, Jae Jun Jeong","doi":"10.1016/j.net.2024.07.041","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.041","url":null,"abstract":"CANDU spent fuels in the Wolsung site have been stored in dry storage systems, such as concrete canisters and modular air-cooled storage system. The primary role of the canister is to ensure the integrity of the fuel during the storage period, which is significantly influenced by temperature. Thus, thermal analysis for the canister's components, especially for fuel cladding, is essential to demonstrate its safety. The thermal analysis has been conducted mainly for predicting the peak cladding temperature (PCT) since high temperature of the fuel can promote oxidation and cracking. As the expiration of storage license approaches, fuel transfer to final disposal should be prepared. This also requires a thermal analysis to predict minimum cladding temperature (MCT), which is related with brittleness. So, it is crucial to accurately predict both PCT and MCT during entire storage period. The cladding temperature is primarily influenced by decay heat and ambient conditions. The lifetime PCT may occur during summer at the beginning of storage, while the lifetime MCT occurs during winter at the end of storage. In this study, we calculated the PCT and MCT during the entire storage period using a realistic thermal analysis model and, subsequently, conducted their uncertainty analysis.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771273","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

Improving tally efficiency and accuracy of multi-group scattering matrix calculations in the Monte Carlo code NECP-MCX 提高蒙特卡罗代码 NECP-MCX 中多组散射矩阵计算的理算效率和精度

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-17 DOI: 10.1016/j.net.2024.07.038

Hongchun Wu, Shuai Qin, Yunzhao Li, Jinkang Shi, Qingming He, Liangzhi Cao

{"title":"Improving tally efficiency and accuracy of multi-group scattering matrix calculations in the Monte Carlo code NECP-MCX","authors":"Hongchun Wu, Shuai Qin, Yunzhao Li, Jinkang Shi, Qingming He, Liangzhi Cao","doi":"10.1016/j.net.2024.07.038","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.038","url":null,"abstract":"Two issues arise in the calculation of the multi-group scattering matrix when employing a continuous-energy Monte Carlo code for generating homogenized multi-group cross-sections. Firstly, the analog estimator is used to evaluate group-to-group elements, which leads to large statistical uncertainty. Secondly, employing the scalar flux as the weighting function in generating the high-order scattering matrix introduces errors in fast reactor calculations. For the first issue, the repeated collision approach and pre-tabulated cross-section approach are adopted to improve the tally efficiency. For the second issue, the average scattering cosine is calculated based on the conservation of the mean square displacement of neutrons, which is then used to correct the first-order self-scattering cross-section. To evaluate the effectiveness of the above approaches, a PWR pin-cell problem and fast reactor core problems are tested. The results demonstrate that: 1) The figure of merit for multi-group scattering matrix calculations was improved by 8–12 times with the pre-tabulated cross-section approach. 2) Biases of were reduced from over 500 pcm to less than 300 pcm when using the corrected self-scattering cross-section. 3) The corrected self-scattering cross-section also yielded higher accuracy for the assembly power calculations, where the maximum biases are reduced from 5 % to 1 %.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141786273","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

Experimental study on practical application of optical fiber sensor (OFS) for high-temperature system 高温系统光纤传感器（OFS）实际应用实验研究

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.025

Byeongyeon Kim, Youngwoong Kim, YunSook Lee, Ki-Ean Nam, Jung Yoon, Yong-Hoon Shin, Hyeonil Kim, Jewhan Lee, BongWan Lee

{"title":"Experimental study on practical application of optical fiber sensor (OFS) for high-temperature system","authors":"Byeongyeon Kim, Youngwoong Kim, YunSook Lee, Ki-Ean Nam, Jung Yoon, Yong-Hoon Shin, Hyeonil Kim, Jewhan Lee, BongWan Lee","doi":"10.1016/j.net.2024.07.025","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.025","url":null,"abstract":"This study explores the application of Raman scattering-based optical fiber sensors (OFSs) in extreme environments, specifically focusing on a loop heater vessel with temperatures ranging from 200 °C to 680 °C. This condition generally covers the advanced reactor designs, such as Sodium-cooled Fast Reactor and High Temperature Reactor. Various optical fiber combinations were employed for temperature measurements, taking into consideration the operating temperature of the target equipment. Two types of OFSs, gold-coated and polyimide-coated, were utilized. Protective tubes made of stainless steel (STS) and carbon were introduced to ensure reliable temperature data collection in high-temperature settings. Results indicate that the STS tube with a gold-coated OFS exhibited the highest consistency and agreement with thermocouple measurements, making it suitable for extreme environments. The study emphasizes the applicability of this system in high-temperature environments, such as liquid metal reactors, high-temperature thermal energy storage system, and hydrogen production system, for environmental monitoring.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612725","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

Numerical study on the size effect on the mixing in 2×1, 3×3 and 5×5 rod bundle subchannels 关于 2×1、3×3 和 5×5 棒束子通道中混合的尺寸效应的数值研究

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.019

Bin Han, Yuanyuan Yin, Xiaoliang zhu, Bao-Wen Yang, Aiguo Liu, Shenghui Liu

{"title":"Numerical study on the size effect on the mixing in 2×1, 3×3 and 5×5 rod bundle subchannels","authors":"Bin Han, Yuanyuan Yin, Xiaoliang zhu, Bao-Wen Yang, Aiguo Liu, Shenghui Liu","doi":"10.1016/j.net.2024.07.019","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.019","url":null,"abstract":"Mixing Vane Grid (MVG) is considered as one of the most important components in the fuel assembly which not only plays the role of supporting the rod bundles but also improves the Critical Heat Flux (CHF) in the reactor core. Modeling and measuring the flow behavior accurately in the rod bundle is the key to understanding and learning complex grid performance in the fuel assembly and will develop high performance MVG. Usually, the fuel assembly in the reactor core consists of 17 × 17 or 16 × 16 rod bundles, it is hardly to use the original MVGs to perform study. The representative smaller prototypical grids are applied. Different bundle sizes are used including 1 × 1, 2 × 1, 3 × 3 and 5 × 5 et al. It is an absolute question of how the smaller size rod bundles are prototypical that could fully reflect the true flow and heat transfer behavior in a reactor core. In this paper, the effect of bundle size on flow and heat transfer is investigated under sizes of 2 × 1, 3 × 3 and 5 × 5. Firstly, the boundary settings in 2 × 1 are studied and the surface averaged secondary flow and local flow at the gap with 5 × 5 results are compared. Then the 3 × 3 and 5 × 5 bundle sizes are compared under subcooled flow. The center subchannels temperature and the void fraction distributions are analyzed. The effect of non-prototypical cold walls on heat transfer is discussed. The study shows that, different bundle sizes will produce different flow phenomena in the rod bundle, the flow pattern may not be the same with the reactor core fuel assembly, the typical bundle size selection should be based on the research purpose.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612796","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

A simplified SP3 NEM solver within a unified formulation for pin-by-pin core multi-group calculations 简化的 SP3 NEM 求解器，采用统一公式进行逐针核心多组计算

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-08 DOI: 10.1016/j.net.2024.07.017

Sicheng Wang, Ser Gi Hong

{"title":"A simplified SP3 NEM solver within a unified formulation for pin-by-pin core multi-group calculations","authors":"Sicheng Wang, Ser Gi Hong","doi":"10.1016/j.net.2024.07.017","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.017","url":null,"abstract":"This study addresses the development and verification of a pin-by-pin core multigroup SP solver CTRP-Clouds that employs NEM (Nodal Expansion Method) and three simplified NEM methods within a unified formulation for simultaneously solving the coupling 0 and 2 SP equations. In this work, the solver using this unified formulation does not only include the original NEM and its simplifications but also the EFEN (Exponential Function Expansion Nodal) method and FDM (Finite Difference Method) for the comprehensive evaluation. Also, the solver was accelerated using CMFD (Coarse Mesh Finite Difference) method and parallelized using OpenMP. The computational efficiency of different solution methods was investigated for the 2D KAIST benchmark problems and their modified one for considering 3D extension. The results showed the simplified NEM with flat leakage approximation gives acceptable accuracies of less than 1.2 % in RMS of pin-power discrepancies of all the cases with 1x1 mesh per pin-cell, with a reduction of 20 % computing time compared to the original NEM. Particularly, the calculation time of flat leakage NEM is comparable to EFEN while the pin-wise accuracy is better. Besides, the simplified NEM with 2-order flux expansion gives substantially improved accuracy in comparison with FDM within comparable computing time.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612793","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

Imbalanced data fault diagnosis method for nuclear power plants based on convolutional variational autoencoding Wasserstein generative adversarial network and random forest 基于卷积变异自动编码 Wasserstein 生成对抗网络和随机森林的核电站不平衡数据故障诊断方法

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-05 DOI: 10.1016/j.net.2024.07.015

Jun Guo, Yulong Wang, Xiang Sun, Shiqiao Liu, Baigang Du

{"title":"Imbalanced data fault diagnosis method for nuclear power plants based on convolutional variational autoencoding Wasserstein generative adversarial network and random forest","authors":"Jun Guo, Yulong Wang, Xiang Sun, Shiqiao Liu, Baigang Du","doi":"10.1016/j.net.2024.07.015","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.015","url":null,"abstract":"Data-driven fault diagnosis techniques are significant for the stable operation of nuclear power plants (NPPs). However, in practical applications, the fault diagnosis of NPPs usually faces imbalance data problems with small fault samples and much redundant data which results in low model training efficiency and poor generalization performance. Thus, this paper proposes a convolutional variational autoencoding gradient-penalty Wasserstein generative adversarial network with random forest (CVGR) to reduce the impact of imbalanced samples on fault diagnosis. Firstly, a feature selection method based on the random forest is used to identify the most relevant measurements and reduce the impact of redundant data on fault diagnosis. Then, variational autoencoding is introduced into gradient-penalty Wasserstein generative adversarial to effectively extract original sample features and generate high-quality samples with high rationality and diversity. In addition, the convolutional neural network is used to extract the features of mixed samples to realize intelligent fault diagnosis. Finally, several experiments based on the Fuqing Unit 2 full-scope simulator under different operating conditions are used to validate the performance of the CVGR in data enhancement and intelligent fault diagnosis. The results show that the proposed method can effectively mitigate the imbalance data problem, which gives insights into intelligent fault diagnosis of NPPs.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141612809","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

Optimizing spent nuclear fuel cask loading for VVER-440 fuel 优化 VVER-440 燃料乏核燃料桶的装载

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-05 DOI: 10.1016/j.net.2024.07.014

M. Lovecký, J. Závorka

引用次数: 0

Hydrodynamics coupled circulating-fuel reactor equations in comoving frame and their analytical solutions for molten salt reactors 移动框架下的水动力学耦合循环燃料反应堆方程及其对熔盐反应堆的解析解

IF 2.7 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-04 DOI: 10.1016/j.net.2024.07.008

Ayhan Yılmazer, Gökhan Pediz

引用次数: 0