Nuclear Engineering and Technology最新文献_第10页

Study of circulating liquid fuel in a 1D critical system with thermal feedback 带有热反馈的一维临界系统中的循环液体燃料研究

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-14 DOI: 10.1016/j.net.2024.07.028

Mathis Caprais , Daniele Tomatis , André Bergeron

{"title":"Study of circulating liquid fuel in a 1D critical system with thermal feedback","authors":"Mathis Caprais , Daniele Tomatis , André Bergeron","doi":"10.1016/j.net.2024.07.028","DOIUrl":"10.1016/j.net.2024.07.028","url":null,"abstract":"<div><div>This research focuses on the description and modeling of a one-dimensional molten salt reactor (MSR), in the presence of thermal feedback. Following the example of previous works, a simple one-dimensional system is proposed, describing a molten salt reactor with a main neutron-multiplying zone called core and a recirculation loop where the salt cools down. Specific attention is paid to the precursors’ drift by modifying the neutron balance equation. Liquid nuclear fuels are characterized by a high volumetric expansion coefficient in comparison to customary solid fuels. Therefore, a strong coupling between neutronics and thermal-hydraulics is expected. As a consequence, a highly negative density coefficient characterizes the thermal feedback on the neutron reactivity. The precursor equation is here inverted analytically and combined with the neutron balance equation to obtain a generalized eigenvalue problem with the neutron flux distribution as the unknown. The balance equations are derived by finite volume integration over a discretized mesh, and the coupling between the two physical models is treated by Picard iterations. The numerical solution is finally extended to time-dependent calculations and compared to an analytical work for a one-dimensional circulating fuel reactor already existing in the literature.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5212-5221"},"PeriodicalIF":2.6,"publicationDate":"2024-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141703088","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Assessing the Phebus FPT-1 experiment: Insights from MELCOR 2.2 and ASYST codes 评估 Phebus FPT-1 试验：从 MELCOR 2.2 和 ASYST 代码中获得的启示

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-14 DOI: 10.1016/j.net.2024.07.021

Peter Mician , Tomas Cerny , Taron Petrosyan , Stepan Foral , Karel Katovsky , Michal Ptacek

引用次数: 0

Effect of thermal neutron flux on borate and silicate glasses: Experimental and theoretical investigation for distribution, decay period, and absorbed dose rate of produced isotopes 热中子通量对硼酸盐和硅酸盐玻璃的影响：对所产生同位素的分布、衰变周期和吸收剂量率的实验和理论研究

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-14 DOI: 10.1016/j.net.2024.07.029

O.L. Tashlykov , K.A. Mahmoud , D.O. Kaskov , T.P. Volozheninov

{"title":"Effect of thermal neutron flux on borate and silicate glasses: Experimental and theoretical investigation for distribution, decay period, and absorbed dose rate of produced isotopes","authors":"O.L. Tashlykov , K.A. Mahmoud , D.O. Kaskov , T.P. Volozheninov","doi":"10.1016/j.net.2024.07.029","DOIUrl":"10.1016/j.net.2024.07.029","url":null,"abstract":"<div><div>The gamma-ray shielding properties for silicate and borate glass samples were enhanced by heavy metal oxides such as Bi<sub>2</sub>O<sub>3</sub>, CdO, and Y<sub>2</sub>O<sub>3</sub>. The current work aims to study experimentally performance of these borate and silicate glasses experimentally when exposed to a flux of thermal neutrons. A nuclear research reactor was utilized to irradiate the fabricated glasses with various doses of thermal neutrons varied between 1.73 and 12.10 MGy. Additionally, the fluency of the thermal neutrons within the dry channel of the IVV-2M reactor containing the fabricated samples varied between 2.27E+17 and 15.86E+17 neutron/cm<sup>2</sup> at various irradiation times between 1 and 7 days. After that, a gamma-ray spectrometer was utilized to detect the activity concentrations from the irradiated glasses as well as identify the new isotopes created within the fabricated glasses. Additionally, the Monte Carlo simulation code was utilized to estimate the absorbed dose from the irradiated glass samples over a time period between 1 and 120 days after the exposure. The study shows almost all of the activities of the irradiated samples decomposed over 80 days after irradiation. The decomposition of the dose rate and activity concentration for the irradiated samples are attributed to the short lifetime of Ba-131 isotope which represents the major radioactive isotope created within the glass sample.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5222-5230"},"PeriodicalIF":2.6,"publicationDate":"2024-07-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141709044","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Al–B4C-(Gd, Gd2O3) composite materials: Synthesis and characterization for neutron shielding applications Al-B4C-(Gd, Gd2O3) 复合材料：中子屏蔽应用的合成与表征

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-11 DOI: 10.1016/j.net.2024.07.027

Yasin Gaylan , Barış Avar

{"title":"Al–B4C-(Gd, Gd2O3) composite materials: Synthesis and characterization for neutron shielding applications","authors":"Yasin Gaylan , Barış Avar","doi":"10.1016/j.net.2024.07.027","DOIUrl":"10.1016/j.net.2024.07.027","url":null,"abstract":"<div><div>In this study, Al–20%B<sub>4</sub>C-x%Gd and Al–20%B<sub>4</sub>C-x%Gd<sub>2</sub>O<sub>3</sub> (x = 1, 3, 5) composite powders were prepared using a high-energy planetary ball milling method to enhance the physical properties of Al–B<sub>4</sub>C neutron shielding composites. The prepared powders were subjected to uniaxial cold compaction at 500 MPa, resulting in cylindrical specimens. Subsequently, the specimens were sintered in a tube furnace at 600 °C for 1 h under an Ar atmosphere to prevent oxidation. The microstructure of the resulting composites was characterized using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). Archimedes density, hardness, and corrosion tests were performed on the compacted samples. Moreover, the composite's thermal and fast neutron absorption rates were calculated using the MCNP6.2 simulation code. The neutron equivalent dose rate was experimentally determined using the Am–Be neutron source. The simulation results demonstrated that the composite materials containing Gd exhibited the highest thermal neutron absorption rate, while those containing Gd<sub>2</sub>O<sub>3</sub> demonstrated the highest fast neutron absorption rate. This research contributes valuable insights into the design and utilization of neutron-absorbing materials with suitable mechanical properties.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5201-5211"},"PeriodicalIF":2.6,"publicationDate":"2024-07-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141697914","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Applicability analysis of induction bending process to P91 piping of PGSFR by high-temperature fatigue test 通过高温疲劳试验分析感应弯曲工艺对 PGSFR P91 管道的适用性

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-11 DOI: 10.1016/j.net.2024.07.026

Tae-Won Na , Nak-Hyun Kim , Chang-Gyu Park , Junehyung Kim , Jong-Bum Kim , Il-Kwon Oh

{"title":"Applicability analysis of induction bending process to P91 piping of PGSFR by high-temperature fatigue test","authors":"Tae-Won Na , Nak-Hyun Kim , Chang-Gyu Park , Junehyung Kim , Jong-Bum Kim , Il-Kwon Oh","doi":"10.1016/j.net.2024.07.026","DOIUrl":"10.1016/j.net.2024.07.026","url":null,"abstract":"<div><div>The application of the induction bending process in pipe fabrication is expanding across industries, significantly reducing leakage by minimizing welded sections in curved pipes. In this study, the applicability of the induction bending process to P91 bent pipes of PGSFR was analyzed, focusing on both material fatigue tests and structural fatigue tests for induction bent pipe at high temperatures. First, both high-cycle and low-cycle fatigue tests on specimens from the bent pipe were carried out at 550 °C to confirm that the fatigue properties meet the ASME Code's fatigue requirements. Second, material constants for a Chaboche combined hardening model were identified by using the material test results and an inelastic finite element analysis of the P91 bent pipe were performed to determine the fatigue test load for structural test effectively. Lastly, a high-temperature fatigue test on the bent pipe structure was performed to assess its structural integrity and post-test non-destructive examination confirmed that no fatigue cracks developed, and thereby affirming the applicability of the P91 bent pipe.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5190-5200"},"PeriodicalIF":2.6,"publicationDate":"2024-07-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141716913","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Impact of choice of fragility approaches on seismic risk quantification of nuclear power plants 选择脆性方法对核电站地震风险量化的影响

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-10 DOI: 10.1016/j.net.2024.07.023

Baris Kasapoglu, Halil Sezen, Tunc Aldemir, Richard Denning

{"title":"Impact of choice of fragility approaches on seismic risk quantification of nuclear power plants","authors":"Baris Kasapoglu, Halil Sezen, Tunc Aldemir, Richard Denning","doi":"10.1016/j.net.2024.07.023","DOIUrl":"10.1016/j.net.2024.07.023","url":null,"abstract":"<div><div>Development and evaluation of seismic fragility of structures and components is crucial in seismic probabilistic risk assessment of nuclear power plants. Simulation-based fragility approaches are a prevailing trend in the literature, while industry-recommended methodologies rely heavily on engineering judgment and deterministic analysis. In this paper, four critical components are selected, modeled and analyzed as a case study. Their fragilities are evaluated using both state-of-the-art fragility methods and code-recommended methods with approximate models. The impact of the choice of fragility approaches on the fragilities of the components and conditional core damage probability of the plant are assessed. The findings reveal that the recommended approaches employed with approximate models have limitations in estimating the median capacity of complex equipment. While there is a notable variance in the treatment of uncertainty among fragility approaches, its influence on core damage probability remains limited unless the component is the primary contributor to core damage.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5154-5174"},"PeriodicalIF":2.6,"publicationDate":"2024-07-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141707876","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","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 2.6 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":"10.1016/j.net.2024.07.025","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5182-5189"},"PeriodicalIF":2.6,"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":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Thermal dynamics aspect identification of loop heat pipe with capillary tube wick using nonlinear autoregressive exogenous neural network 利用非线性自回归外源神经网络识别带毛细管芯的环形热管的热动力学特性

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.022

Yoyok Dwi Setyo Pambudi , Giarno , Sumantri Hatmoko , Anhar Riza Antariksawan , Mukhsinun Hadi Kusuma

{"title":"Thermal dynamics aspect identification of loop heat pipe with capillary tube wick using nonlinear autoregressive exogenous neural network","authors":"Yoyok Dwi Setyo Pambudi , Giarno , Sumantri Hatmoko , Anhar Riza Antariksawan , Mukhsinun Hadi Kusuma","doi":"10.1016/j.net.2024.07.022","DOIUrl":"10.1016/j.net.2024.07.022","url":null,"abstract":"<div><div>The loop heat pipe (LHP) has the potential to be used as a passive cooling system in small modular reactors. The research objective is to study the thermal dynamics of LHP with capillary tube wick using a non-linear autoregressive exogenous (NARX) based on a neural network. The neural network identification of LHP with capillary tube wick was carried out on the MATLAB platform. The experiment data obtained is used to identify the neural network of LHP with capillary tube wick. The temperature of the water as an evaporator heat source was varied at 60, 70, 80, and 90 °C. The LHP was charged with demineralized water with a filling ratio of 100 %. The air as a coolant in condenser section was blown at velocity of 2.5 m/s. The LHP was vacuumed with an initial pressure of 2690 Pa. The result confirmed that NARX based on the neural network model can predict the temperature of the condenser section with a given input set under the steady-state and transient conditions. The coefficient of determination is higher than 0.998 and Mean Square Error (MSE) is below 0.0082. The result obtained shows that the NARX neural network model can predict thermal dynamics phenomena in LHP quickly and precisely.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5145-5153"},"PeriodicalIF":2.6,"publicationDate":"2024-07-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141708888","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Methodology for estimating the contribution of forest fires in loss of offsite power events 估算森林火灾在异地电力损失事件中所占比例的方法

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.018

Joonseok Lim, Seungsu Han, Hyungdae Kim, Gyunyoung Heo

引用次数: 0

Improvement of macroscopic turbulence model for subchannel analysis in the rod bundle array 改进用于杆束阵列子通道分析的宏观湍流模型

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2024-07-09 DOI: 10.1016/j.net.2024.07.020

Seok Kim, Jee Min Yoo, Sang-Ki Moon

{"title":"Improvement of macroscopic turbulence model for subchannel analysis in the rod bundle array","authors":"Seok Kim, Jee Min Yoo, Sang-Ki Moon","doi":"10.1016/j.net.2024.07.020","DOIUrl":"10.1016/j.net.2024.07.020","url":null,"abstract":"<div><div>The PRIUS program was established to generate an experimental database for the 6 × 4 and 12 × 6 rod bundle geometry. The database will be used to address the subchannel and CFD code analysis required for modeling and validation. This is necessary because Small Break Loss of Coolant Accident (SBLOCA) and Intermediate Break Loss of Coolant Accident (IBLOCA) present three-dimensional phenomena in the core due to the radial power profile, crossflow, and diffusion-dispersion. Therefore, specific experimental programs are required, especially during core reflooding, to investigate the large-scale three-dimensional effects. However, validating each sensitive model of the code separately in the presence of 3D effects is not possible due to the inability to implement instrumentation at high pressure and temperature steam-water flow conditions. The PRIUS test program uses a single-phase flow test to simulate a non-homogeneous velocity distribution and provide information on crossflow with radial mixing effects between subchannels. The CUPID code, which uses a macroscopic turbulence model, has been validated using the PRIUS-II experimental database. Existing macroscopic turbulence models were also validated for their prediction capabilities with different inlet flow conditions. However, the validation revealed significant errors in the shear region between subchannels. An improved macroscopic turbulence model showed promising results in predicting turbulence kinetic energy in porous media analysis.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"56 12","pages":"Pages 5118-5135"},"PeriodicalIF":2.6,"publicationDate":"2024-07-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"141691763","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0