Pub Date : 2024-07-23DOI: 10.1016/j.net.2024.07.045
Xicheng Wang, Govatsa Acharya, Dmitry Grishchenko, Pavel Kudinov
Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the ‘Unit cell’ approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.
{"title":"CFD Simulation of Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression Pool","authors":"Xicheng Wang, Govatsa Acharya, Dmitry Grishchenko, Pavel Kudinov","doi":"10.1016/j.net.2024.07.045","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.045","url":null,"abstract":"Boiling Water Reactor (BWR) employs the Pressure Suppression Pool (PSP) as a heat sink to prevent overpressure of the reactor vessel and containment. Steam can be injected into the PSP through spargers in normal and accident conditions and through blowdown pipes in case of a loss of coolant accident (LOCA). There is a safety limit on the maximum PSP temperature at which such steam injection might cause dynamic loads on the containment structures. The performance of the pool can be affected if thermal stratification is developed when temperature of the hot layer grows rapidly while cold layer remains inactive. Simulation of pool behavior during realistic accident scenarios requires validated models that can sufficiently address the interaction between phenomena, safety systems and operational procedures. Direct modeling of steam injection into a water pool in long-term transients is computationally expensive due to the need to resolve simultaneously the smallest space and time scales of individual steam bubbles and the scales of the whole PSP. To enable PSP analysis for practical purposes, Effective Heat source and Effective Momentum source (EHS/EMS) models have been proposed that avoid the need to resolve steam-water interface. This paper aims to implement mechanistic approaches previously developed by authors for the simulation of transient thermal stratification and mixing phenomena induced by steam injection through spargers in a Nordic BWR PSP. The latest version of the EHS/EMS models using the ‘Unit cell’ approach has been validated against integral effect pool tests and applied to plant simulations. Several scenarios with boundary conditions corresponding to postulated accident sequences were simulated to investigate the possibility of stratification development and the effects of activation of different systems (e.g., blowdown pipes, high momentum nozzle) on the pool behavior.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141771274","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}
Pub Date : 2024-07-22DOI: 10.1016/j.net.2024.07.044
Rashed MD. Sardar, Akhmed M. Baisov
The evaluation of available Look-Up Tables for prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water with the aim of their further use in computational analyses of various Fuel Assembles of Supercritical Water-Cooled Reactors is made. The comparison between the calculations based on Look-Up Tables with the values from empirical correlations and experimental data for smooth and wire-wrapped rod bundles was presented. The obtained results showed that Look-Up Table of the University of Ottawa, which was created to describe improved and deteriorated heat transfer regimes in round tubes, allows describing available data points with 30% of the mean square deviation. It is noted that the presence of wire intensifies heat transfer exchange near pseudocritical temperature region but existing versions of Look-Up Tables cannot take into account this effect. Nevertheless, there is potential for further improvement in predicting the heat transfer coefficient using Look-Up Table by introducing additional correction factors.
{"title":"ASSESSMENT OF LOOK-UP TABLES FOR THE PREDICTION OF HEAT TRANSFER COEFFICIENT DISTRIBUTION IN ROD BUNDLES COOLED BY SUPERCRITICAL WATER","authors":"Rashed MD. Sardar, Akhmed M. Baisov","doi":"10.1016/j.net.2024.07.044","DOIUrl":"https://doi.org/10.1016/j.net.2024.07.044","url":null,"abstract":"The evaluation of available Look-Up Tables for prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water with the aim of their further use in computational analyses of various Fuel Assembles of Supercritical Water-Cooled Reactors is made. The comparison between the calculations based on Look-Up Tables with the values from empirical correlations and experimental data for smooth and wire-wrapped rod bundles was presented. The obtained results showed that Look-Up Table of the University of Ottawa, which was created to describe improved and deteriorated heat transfer regimes in round tubes, allows describing available data points with 30% of the mean square deviation. It is noted that the presence of wire intensifies heat transfer exchange near pseudocritical temperature region but existing versions of Look-Up Tables cannot take into account this effect. Nevertheless, there is potential for further improvement in predicting the heat transfer coefficient using Look-Up Table by introducing additional correction factors.","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":"141771272","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}
Pub Date : 2024-07-22DOI: 10.1016/j.net.2024.07.046
Fan Miao, Bin Zhang, Tianci Xie, Hao Yang, Jianqiang Shan
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.
{"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}
Pub Date : 2024-07-20DOI: 10.1016/j.net.2024.07.042
Xiaoliang Zou, Yanting Sun, Qiusun Zeng, Xiaojian Wen, Xiaogang Cao, Xi Huang, Yibao Liu
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.
用热管冷却的空间核反应堆已成为未来空间任务和深空探测任务的首选。为实现全球核不扩散的目标，低浓缩铀（LEU）的使用得到了推广。本研究基于热管-分段式热电模块转换器（HP-STMCs），提出了一种使用 LEU 的锂热管冷却空间反应堆（HP-LEU），并增加了慢化剂。HP-LEU 采用氢化钇（YH）作为慢化剂，19.9% 的浓缩氮化铀（UN）作为燃料。对 HP-LEU 反应堆进行了中子分析，结果表明 HP-LEU 的寿命超过 12 年。反应堆采用了两套控制系统，并证明其具有独立调节和关闭反应堆的能力。总温度反应系数始终为负值，表明 HP-LEU 反应堆在运行期间本质上是安全的。在正常运行期间，材料的温度都是可以接受的。这项研究可作为使用 LEU 的锂热管冷却空间反应堆的参考。
{"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}
Pub Date : 2024-07-18DOI: 10.1016/j.net.2024.07.041
Tae Gang Lee, Taehyung Na, Byongjo Yun, Jae Jun Jeong
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.
{"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}
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 %.
{"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}
Pub Date : 2024-07-09DOI: 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
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.
本研究探讨了基于拉曼散射的光纤传感器（OFS）在极端环境中的应用，尤其侧重于温度范围为 200 °C 至 680 °C 的循环加热器容器。这种条件一般涵盖先进的反应堆设计，如钠冷快堆和高温反应堆。考虑到目标设备的工作温度，采用了各种光纤组合进行温度测量。使用了两种类型的 OFS：金涂层和聚酰亚胺涂层。为了确保在高温环境下可靠地收集温度数据，还采用了不锈钢（STS）和碳纤维制成的保护管。结果表明，带有金涂层 OFS 的 STS 管与热电偶测量结果的一致性和一致性最高，因此适用于极端环境。这项研究强调了该系统在高温环境中的适用性，如液态金属反应堆、高温热能存储系统和制氢系统的环境监测。
{"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}
Pub Date : 2024-07-09DOI: 10.1016/j.net.2024.07.019
Bin Han, Yuanyuan Yin, Xiaoliang zhu, Bao-Wen Yang, Aiguo Liu, Shenghui Liu
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.
{"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}
Pub Date : 2024-07-08DOI: 10.1016/j.net.2024.07.017
Sicheng Wang, Ser Gi Hong
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.
本研究针对逐针核心多组 SP 求解器 CTRP-Clouds 的开发和验证，该求解器在统一公式中采用了 NEM（节点扩展法）和三种简化 NEM 方法，可同时求解耦合 0 和耦合 2 SP 方程。在这项工作中，使用这种统一公式的求解器不仅包括原始 NEM 及其简化方法，还包括用于综合评估的 EFEN（指数函数展开节点法）和 FDM（有限差分法）。此外，还使用 CMFD（粗网格有限差分）方法加速了求解器，并使用 OpenMP 进行了并行化。针对二维 KAIST 基准问题和考虑三维扩展的修正问题，研究了不同求解方法的计算效率。结果表明，在每个引脚单元采用 1x1 网格的情况下，采用平面泄漏近似的简化 NEM 在所有情况下的引脚功率差异均方根有效值（RMS）小于 1.2 %，计算时间比原始 NEM 减少了 20 %。特别是，平面漏电 NEM 的计算时间与 EFEN 相当，而引脚精度更高。此外，在计算时间相当的情况下，采用二阶通量扩展的简化 NEM 与 FDM 相比，精度大幅提高。
{"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}
Pub Date : 2024-07-05DOI: 10.1016/j.net.2024.07.015
Jun Guo, Yulong Wang, Xiang Sun, Shiqiao Liu, Baigang Du
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.
{"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}
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