Pub Date : 2024-09-01DOI: 10.1016/j.net.2024.08.062
Jinkyun Park, Wasin Vechgama, Seung Ki Shin
Operating teams consisting of several team members still play a critical role in coping with off-normal conditions in socio-technical systems. Thus, various kinds of human reliability analysis methods have been suggested based on the consideration of diverse performance shaping factors that can affect the performance of team members. Unfortunately, since multiple performance shaping factors can vary across operating teams (i.e., crew-to-crew variability), it is crucial to figure out how to visualize this variability in a systematic way. In this regard, comparing the cultural characteristics of operating teams with their performance would be a good starting point. This study investigates how cultural characteristics can be correlated with the occurrence of unsafe acts based on empirical data collected from operating teams working in the main control room of Korean domestic nuclear power plants. The cultural characteristics of the operating teams were visualized using five Hofstede's cultural indices and compared with the number of unsafe acts observed from simulated off-normal conditions. As a result, a statistically significant correlation is found between the occurrence of unsafe acts and one of the Hofstede's indices. From this finding, it is expected that a relevant probe to scrutinize crew-to-crew variability could be soundly determined in future works.
{"title":"Identifying a probe to visualize the variability of operating teams for supporting the human reliability analysis of nuclear power plants: An explanatory study","authors":"Jinkyun Park, Wasin Vechgama, Seung Ki Shin","doi":"10.1016/j.net.2024.08.062","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.062","url":null,"abstract":"Operating teams consisting of several team members still play a critical role in coping with off-normal conditions in socio-technical systems. Thus, various kinds of human reliability analysis methods have been suggested based on the consideration of diverse performance shaping factors that can affect the performance of team members. Unfortunately, since multiple performance shaping factors can vary across operating teams (i.e., crew-to-crew variability), it is crucial to figure out how to visualize this variability in a systematic way. In this regard, comparing the cultural characteristics of operating teams with their performance would be a good starting point. This study investigates how cultural characteristics can be correlated with the occurrence of unsafe acts based on empirical data collected from operating teams working in the main control room of Korean domestic nuclear power plants. The cultural characteristics of the operating teams were visualized using five Hofstede's cultural indices and compared with the number of unsafe acts observed from simulated off-normal conditions. As a result, a statistically significant correlation is found between the occurrence of unsafe acts and one of the Hofstede's indices. From this finding, it is expected that a relevant probe to scrutinize crew-to-crew variability could be soundly determined in future works.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207218","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 : 2024-09-01DOI: 10.1016/j.net.2024.08.064
Paul Hurley, Yang Liu, Tomasz Kozlowski, Juliana P. Duarte
Density wave oscillation (DWO) is an important safety concern for boiling water reactors (BWR) due to their high void fraction in the core. Power extensions to existing reactors such as the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) lead to increase susceptibility of DWO-type instability following an anticipated transient without scram (ATWS). Experiments performed at the Karlstein Thermal Hydraulic Test Facility (KATHY) have reproduced the reactivity feedback mechanism in BWRs under ATWS conditions. Using a neutronics module simulator, the KATHY facility was able to provide data on the effect of different neutronic parameters on DWO onset. This paper serves to assess the capability of the thermal-hydraulics code TRACE V5P7 for simulating DWO onset and development under natural circulation with neutronic feedback. A model of the KATHY natural circulation facility is created in TRACE and a reactivity feedback mechanism is implemented using a manual control scheme to simulate the parametric effects provided by the tests. This comparison allows for an assessment of the TRACE code as well as a better understanding of the instability mechanisms and behavior under the given conditions.
{"title":"TRACE assessment of density wave instability onset with void reactivity feedback under natural circulation","authors":"Paul Hurley, Yang Liu, Tomasz Kozlowski, Juliana P. Duarte","doi":"10.1016/j.net.2024.08.064","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.064","url":null,"abstract":"Density wave oscillation (DWO) is an important safety concern for boiling water reactors (BWR) due to their high void fraction in the core. Power extensions to existing reactors such as the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) lead to increase susceptibility of DWO-type instability following an anticipated transient without scram (ATWS). Experiments performed at the Karlstein Thermal Hydraulic Test Facility (KATHY) have reproduced the reactivity feedback mechanism in BWRs under ATWS conditions. Using a neutronics module simulator, the KATHY facility was able to provide data on the effect of different neutronic parameters on DWO onset. This paper serves to assess the capability of the thermal-hydraulics code TRACE V5P7 for simulating DWO onset and development under natural circulation with neutronic feedback. A model of the KATHY natural circulation facility is created in TRACE and a reactivity feedback mechanism is implemented using a manual control scheme to simulate the parametric effects provided by the tests. This comparison allows for an assessment of the TRACE code as well as a better understanding of the instability mechanisms and behavior under the given conditions.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207217","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}
In order to support the verification of neutronics/thermal-hydraulics coupling calculation method or simulation codes at the fine mesh for plate-type pressurized water reactor (PWR) with high parameters (which the ratio of power to mass flow rate is greater than 235 kW/kg, the core outlet enthalpy exceeds 1500 kJ/kg), a set of coupling calculation of plate-type PWR based on high parameters (COPHP) is design by Nuclear Power Institute of China (NPIC). Multiple industry research teams, including Tsinghua University, Xi'an Jiaotong University, and Huazhong University of Science and Technology, participated in the production of COPHP benchmark. This article provides a detailed explanation of the completed assembly design and modeling calculations, and provides the , fine mesh relative power distribution, and deviation results calculated for 10 conditions of 6 assemblies using RMC, OpenMC, and KYLIN V2 software. The results show that: 1) when using the same cross-sectional library, the OpenMC calculation results are in good agreement with the RMC results. Taking ENDF/B-VII.1 as an example, the maximum deviation of in the entire burnup process of all assemblies is −157pcm, the maximum deviation of relative power is −1.13 %, and the maximum power weight error(PWE) is 0.226 %. 2) Compared between KYLIN V2 and the RMC by using ENDF/B-VII.1, the maximum deviation of is −468pcm, the maximum deviation of relative power is −1.49 %, and the maximum PWE is 0.316 % when calculating the standard assemblies with all control rod out(ARO) condition and two burnable poison assemblies. The maximum deviation of is −795pcm, the maximum deviation of relative power is 1.81 %, and the maximum PWE is 0.369 % when calculating the standard assemblies with all control rod inserted(ARI) conditions.
{"title":"Assemblies design and modeling analysis of a new fine mesh neutronics/thermal-hydraulics coupling benchmark for plate-type PWR core","authors":"Zhigang Li, Wei Lu, Shanfang Huang, Xiafeng Zhou, Yingwei Wu, Bangyang Xia, Junji Chen, Tao He, Guodong Liu, Yangyu Zhong, Zhiying Yue","doi":"10.1016/j.net.2024.08.058","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.058","url":null,"abstract":"In order to support the verification of neutronics/thermal-hydraulics coupling calculation method or simulation codes at the fine mesh for plate-type pressurized water reactor (PWR) with high parameters (which the ratio of power to mass flow rate is greater than 235 kW/kg, the core outlet enthalpy exceeds 1500 kJ/kg), a set of coupling calculation of plate-type PWR based on high parameters (COPHP) is design by Nuclear Power Institute of China (NPIC). Multiple industry research teams, including Tsinghua University, Xi'an Jiaotong University, and Huazhong University of Science and Technology, participated in the production of COPHP benchmark. This article provides a detailed explanation of the completed assembly design and modeling calculations, and provides the , fine mesh relative power distribution, and deviation results calculated for 10 conditions of 6 assemblies using RMC, OpenMC, and KYLIN V2 software. The results show that: 1) when using the same cross-sectional library, the OpenMC calculation results are in good agreement with the RMC results. Taking ENDF/B-VII.1 as an example, the maximum deviation of in the entire burnup process of all assemblies is −157pcm, the maximum deviation of relative power is −1.13 %, and the maximum power weight error(PWE) is 0.226 %. 2) Compared between KYLIN V2 and the RMC by using ENDF/B-VII.1, the maximum deviation of is −468pcm, the maximum deviation of relative power is −1.49 %, and the maximum PWE is 0.316 % when calculating the standard assemblies with all control rod out(ARO) condition and two burnable poison assemblies. The maximum deviation of is −795pcm, the maximum deviation of relative power is 1.81 %, and the maximum PWE is 0.369 % when calculating the standard assemblies with all control rod inserted(ARI) conditions.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207220","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}
At present, the CRAFT (Comprehensive Research Facility for Fusion Technology) dual-driver negative ion source for neutral beam injection applications has made progress. In order to solve the problem of heat removal function in large-area extraction grid, the multi-physics simulation, manufacturing and testing are carried out. Two simulation conditions before Cs seeding (J∼350 A/m, J-∼50 A/m) and after sufficient Cs conditioning (J∼150 A/m, J-∼250 A/m) were considered to evaluate the power load on EG, which peak power density were 11 and 6 MW/m respectively and were concentrated on locate near the two aperture rows. This paper mainly studies three designs of minor cooling channels (double bending channels, double straight channels, and single straight channel). First, the grid temperature, thermal stress and thermal deformation are compared with an inlet water rate of 1.21 kg/s at 35 °C, and the channel forming methods and welding types mechanisms are compared. The simulation results show that the thermos-mechanical property of double bending channels is the best, and that of double straight channels is slightly worse. In terms of processing and manufacturing, the yield of double bending channels is lower than double straight channels, and it is easy to leak during the experiment.
{"title":"Optimization of a large-area grid electrode for negative ion source in fusion neutral beam injector","authors":"Jinhong Wang, Yan Wang, Yuwen Yang, Yuming Gu, Yahong Xie, Lizhen Liang, Jianglong Wei","doi":"10.1016/j.net.2024.08.051","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.051","url":null,"abstract":"At present, the CRAFT (Comprehensive Research Facility for Fusion Technology) dual-driver negative ion source for neutral beam injection applications has made progress. In order to solve the problem of heat removal function in large-area extraction grid, the multi-physics simulation, manufacturing and testing are carried out. Two simulation conditions before Cs seeding (J∼350 A/m, J-∼50 A/m) and after sufficient Cs conditioning (J∼150 A/m, J-∼250 A/m) were considered to evaluate the power load on EG, which peak power density were 11 and 6 MW/m respectively and were concentrated on locate near the two aperture rows. This paper mainly studies three designs of minor cooling channels (double bending channels, double straight channels, and single straight channel). First, the grid temperature, thermal stress and thermal deformation are compared with an inlet water rate of 1.21 kg/s at 35 °C, and the channel forming methods and welding types mechanisms are compared. The simulation results show that the thermos-mechanical property of double bending channels is the best, and that of double straight channels is slightly worse. In terms of processing and manufacturing, the yield of double bending channels is lower than double straight channels, and it is easy to leak during the experiment.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207222","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 : 2024-08-30DOI: 10.1016/j.net.2024.08.057
Jong-Gu Kwak, J.H. Jeong, S.C. Hong
KSTAR shows the long pulse capability based on superconducting magnet and NBI plays a crucial role in sustaining the plasma via plasma heating and current driving. So, the accurate NBI power measurement transferred to KSTAR plasma is important for the analysis of plasma transport as well as plasma performance parameters. In a long pulse operation, the water coolant temperature is at the steady state condition in longer pulse more than 20s and the coupled neutral beam power to KSTAR plasma during the tokamak experiments is rechecked by using water flow calorimetric method after the experiment. The effect of beam duct scraper loss which was not considered at the neutral beam power calibration process is less than 5 % in terms of neutral beam power. However, in long pulse operation of NBI in KSTAR experiments, high strength of stray PF affects the beam path, the neutral beam power registered on mds + using beam current method is over estimated and it is calculated up to a few percent in terms of neutral beam power using calorimetric method. Therefore, it is necessary to consider the beam power variation by PF effect to interpret the plasma performance degradation in long pulse operation. Lastly. when the ion source tun on or turn off in condition other ion sources are operated, the beam transmission power is also affected because of sharing the beam box for ion sources so that the careful power estimation is necessary for such kind of beam power modulation experiments in KSTAR.
{"title":"Estimation of neutral beam injector power transferred to KSTAR plasmas","authors":"Jong-Gu Kwak, J.H. Jeong, S.C. Hong","doi":"10.1016/j.net.2024.08.057","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.057","url":null,"abstract":"KSTAR shows the long pulse capability based on superconducting magnet and NBI plays a crucial role in sustaining the plasma via plasma heating and current driving. So, the accurate NBI power measurement transferred to KSTAR plasma is important for the analysis of plasma transport as well as plasma performance parameters. In a long pulse operation, the water coolant temperature is at the steady state condition in longer pulse more than 20s and the coupled neutral beam power to KSTAR plasma during the tokamak experiments is rechecked by using water flow calorimetric method after the experiment. The effect of beam duct scraper loss which was not considered at the neutral beam power calibration process is less than 5 % in terms of neutral beam power. However, in long pulse operation of NBI in KSTAR experiments, high strength of stray PF affects the beam path, the neutral beam power registered on mds + using beam current method is over estimated and it is calculated up to a few percent in terms of neutral beam power using calorimetric method. Therefore, it is necessary to consider the beam power variation by PF effect to interpret the plasma performance degradation in long pulse operation. Lastly. when the ion source tun on or turn off in condition other ion sources are operated, the beam transmission power is also affected because of sharing the beam box for ion sources so that the careful power estimation is necessary for such kind of beam power modulation experiments in KSTAR.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207221","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}
In order to improve the quality of neutron data for U in CENDL, considering the impact of new measurements, a brand-new evaluation of the complete set of neutron induced U reaction data up to 20 MeV has been performed. Important reactions, such as (n,tot), (n,γ), (n,f), (n,2n) and (n,3n) reaction cross sections and average number of fission neutrons have been evaluated based on experimental data analysis. Also, using existing optical model potential parameters, new theoretical calculations based on Hauser-Feshbach and pre-equilibrium model have been carried out. Resonance parameters and prompt fission neutron spectrum from ENDF/B-VIII.0 have been adopted. Guided by integral benchmark, (n,inl), (n,γ) and (n,f) reaction cross sections have undergone multiple adjustments, leading to a noticeable improvement in data quality, as indicated by final benchmark results.
{"title":"Evaluations and calculations of neutron reactions on 238U up to 20 MeV","authors":"Yue Zhang, Ruirui Xu, Yuan Tian, Haicheng Wu, Xichao Ruan, Yinlu Han, Huanyu Zhang, Ping Liu, Xi Tao, Zhi Zhang, Changlin Lan, Xiaodong Sun, Yongli Jin, Nengchuan Shu, Jimin Wang, Xiaolong Huang, Zhigang Ge","doi":"10.1016/j.net.2024.08.059","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.059","url":null,"abstract":"In order to improve the quality of neutron data for U in CENDL, considering the impact of new measurements, a brand-new evaluation of the complete set of neutron induced U reaction data up to 20 MeV has been performed. Important reactions, such as (n,tot), (n,γ), (n,f), (n,2n) and (n,3n) reaction cross sections and average number of fission neutrons have been evaluated based on experimental data analysis. Also, using existing optical model potential parameters, new theoretical calculations based on Hauser-Feshbach and pre-equilibrium model have been carried out. Resonance parameters and prompt fission neutron spectrum from ENDF/B-VIII.0 have been adopted. Guided by integral benchmark, (n,inl), (n,γ) and (n,f) reaction cross sections have undergone multiple adjustments, leading to a noticeable improvement in data quality, as indicated by final benchmark results.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207219","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 : 2024-08-28DOI: 10.1016/j.net.2024.08.053
Jae Sung Shin, Joonsoo Ock, Sungyeol Choi
We conducted laser cutting studies on zirconium alloys, specifically Zircaloy-2 and Zr-2.5%Nb alloy, which are used as constituent materials in the nuclear fuel channel of a pressurized heavy water reactor. The study measured the maximum cutting speed, amount of secondary emissions, and aerosol characteristics for each material using 10 mm thick plate specimens. The cutting performance of the zirconium alloys was similar to each other. At a laser power of 1–5 kW, the maximum cutting speed ranged from 750 to 1900 mm/min, and the amount of secondary emissions per length ranged from 32 to 53 g/m. Compared to 304 L stainless steel, the maximum cutting speed was 1.7–1.9 times higher, and the amount of secondary emissions was about 60–70 % of that of stainless steel. Analyzing the physical properties of aerosols, both zirconium alloys generated larger particles with a count median aerodynamic diameter of 0.25 μm, which is approximately 15–17 % larger than that of 304 L stainless steel.
{"title":"Laser cutting study of zirconium alloys for nuclear decommissioning","authors":"Jae Sung Shin, Joonsoo Ock, Sungyeol Choi","doi":"10.1016/j.net.2024.08.053","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.053","url":null,"abstract":"We conducted laser cutting studies on zirconium alloys, specifically Zircaloy-2 and Zr-2.5%Nb alloy, which are used as constituent materials in the nuclear fuel channel of a pressurized heavy water reactor. The study measured the maximum cutting speed, amount of secondary emissions, and aerosol characteristics for each material using 10 mm thick plate specimens. The cutting performance of the zirconium alloys was similar to each other. At a laser power of 1–5 kW, the maximum cutting speed ranged from 750 to 1900 mm/min, and the amount of secondary emissions per length ranged from 32 to 53 g/m. Compared to 304 L stainless steel, the maximum cutting speed was 1.7–1.9 times higher, and the amount of secondary emissions was about 60–70 % of that of stainless steel. Analyzing the physical properties of aerosols, both zirconium alloys generated larger particles with a count median aerodynamic diameter of 0.25 μm, which is approximately 15–17 % larger than that of 304 L stainless steel.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207226","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 : 2024-08-28DOI: 10.1016/j.net.2024.08.054
Amruta Pasarkar, S. Balaguru
In sodium-cooled fast reactors (SFR), 316L(N) SS grid plate is hardfaced with Ni-Cr-B-Si alloy to achieve higher wear resistance. Tensile and fatigue forces are acting at the interface between substrate and deposit due to different thermal expansion coefficients of those two materials, which can cause cracking of deposit and fracture during operation. Thus, it is very important to consider appropriate hardfacing method which can provide higher tensile and fatigue strength to avoid cracking/debonding at the interface. To find a solution to this problem, two hardfacing techniques, namely Gas Tungsten Arc (GTA) and Laser cladding (LC), are taken into consideration. Hardfaced specimens are prepared using each process on which tensile and high cycle fatigue tests are conducted. From the experimental testing, stress-strain and S-N curves are generated to predict the tensile and fatigue behaviour of specimens. Fractographic studies are conducted at fractured surfaces to understand the fatigue crack nucleation and propagation characteristics. The experimental results for both processes are compared. Tensile and fatigue strength of LC specimens are ∼11 % and ∼17 % less than those of GTA specimens due to its higher brittleness. Thus, GTA process is recommended as the efficient hardfacing process for grid plate of SFR.
{"title":"Comparative study on tensile and high cycle fatigue behaviour of 316L(N) SS hardfaced with Ni-Cr-B-Si alloy by GTA and laser cladding processes","authors":"Amruta Pasarkar, S. Balaguru","doi":"10.1016/j.net.2024.08.054","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.054","url":null,"abstract":"In sodium-cooled fast reactors (SFR), 316L(N) SS grid plate is hardfaced with Ni-Cr-B-Si alloy to achieve higher wear resistance. Tensile and fatigue forces are acting at the interface between substrate and deposit due to different thermal expansion coefficients of those two materials, which can cause cracking of deposit and fracture during operation. Thus, it is very important to consider appropriate hardfacing method which can provide higher tensile and fatigue strength to avoid cracking/debonding at the interface. To find a solution to this problem, two hardfacing techniques, namely Gas Tungsten Arc (GTA) and Laser cladding (LC), are taken into consideration. Hardfaced specimens are prepared using each process on which tensile and high cycle fatigue tests are conducted. From the experimental testing, stress-strain and S-N curves are generated to predict the tensile and fatigue behaviour of specimens. Fractographic studies are conducted at fractured surfaces to understand the fatigue crack nucleation and propagation characteristics. The experimental results for both processes are compared. Tensile and fatigue strength of LC specimens are ∼11 % and ∼17 % less than those of GTA specimens due to its higher brittleness. Thus, GTA process is recommended as the efficient hardfacing process for grid plate of SFR.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207225","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 : 2024-08-28DOI: 10.1016/j.net.2024.08.052
Almog Biton, Evgeny Rabinovich, Erez Gilad
This work presents an experimental study of five threshold velocities for air-water flow in three different vertical channels. The measured threshold velocities included onset flooding (OF), end flooding (EF), onset deflooding (OD), end deflooding (ED), and minimum pressure (MP) velocities. The experimental system includes a transparent vertical tube of 52.5 mm inner diameter and 1500 mm length. The test channel can be easily changed from a tube to an annular shape by inserting a cylindrical test element. A counter-current or concurrent upward flow was achieved by blowing air upward from the channel's bottom and flowing water from its top. The threshold velocities were determined by analyzing the pressure drop versus air superficial velocity. Findings revealed evident hysteresis between the end flooding and onset deflooding velocities. In contrast, the end deflooding and onset flooding velocities were found to be identical. The end flooding velocity was indifferent to the water's superficial velocity for all three channel geometries. A generalized gas-liquid flow state diagram for vertical channels is developed based on the present empirical analysis for different threshold velocities.
{"title":"Experimental investigation of threshold velocities for air-water two-phase flow in a vertical tube and annular channels","authors":"Almog Biton, Evgeny Rabinovich, Erez Gilad","doi":"10.1016/j.net.2024.08.052","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.052","url":null,"abstract":"This work presents an experimental study of five threshold velocities for air-water flow in three different vertical channels. The measured threshold velocities included onset flooding (OF), end flooding (EF), onset deflooding (OD), end deflooding (ED), and minimum pressure (MP) velocities. The experimental system includes a transparent vertical tube of 52.5 mm inner diameter and 1500 mm length. The test channel can be easily changed from a tube to an annular shape by inserting a cylindrical test element. A counter-current or concurrent upward flow was achieved by blowing air upward from the channel's bottom and flowing water from its top. The threshold velocities were determined by analyzing the pressure drop versus air superficial velocity. Findings revealed evident hysteresis between the end flooding and onset deflooding velocities. In contrast, the end deflooding and onset flooding velocities were found to be identical. The end flooding velocity was indifferent to the water's superficial velocity for all three channel geometries. A generalized gas-liquid flow state diagram for vertical channels is developed based on the present empirical analysis for different threshold velocities.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207224","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 : 2024-08-27DOI: 10.1016/j.net.2024.08.049
Shuiqiang Duan, Minggang Li, Jiaming Li, Tongxi Li, Changhua Nie, Zumao Yang, Jun Hu
The accurate measurement of the water level on the secondary side of the steam generator (SG) plays a crucial role in the safe and stable operation of the primary and secondary circuits of the nuclear power plant. In order to study the feasibility of applying the magnetostrictive liquid level gauge to the water level measurement of SG, the measurement results of the magnetostrictive liquid level gauge under the steady and transient conditions were obtained and compared with the traditional differential pressure liquid level gauge. The results indicated that the magnetostrictive liquid level gauge had good measurement accuracy under the cold-steady condition, thermal-steady condition and transient pressurization condition. The design parameters of float and operating parameters were necessary to correct the measurement results of magnetostrictive water level gauge. The measurement results under the transient depressurization condition were affected by the fluid movement in the container, resulting in a decrease in measurement accuracy. The results indicated that the magnetostrictive liquid level gauge had the potential to be used for water level measurement of SG in nuclear power plant.
{"title":"Research on magnetostrictive liquid level gauge for water level measurement of steam generator","authors":"Shuiqiang Duan, Minggang Li, Jiaming Li, Tongxi Li, Changhua Nie, Zumao Yang, Jun Hu","doi":"10.1016/j.net.2024.08.049","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.049","url":null,"abstract":"The accurate measurement of the water level on the secondary side of the steam generator (SG) plays a crucial role in the safe and stable operation of the primary and secondary circuits of the nuclear power plant. In order to study the feasibility of applying the magnetostrictive liquid level gauge to the water level measurement of SG, the measurement results of the magnetostrictive liquid level gauge under the steady and transient conditions were obtained and compared with the traditional differential pressure liquid level gauge. The results indicated that the magnetostrictive liquid level gauge had good measurement accuracy under the cold-steady condition, thermal-steady condition and transient pressurization condition. The design parameters of float and operating parameters were necessary to correct the measurement results of magnetostrictive water level gauge. The measurement results under the transient depressurization condition were affected by the fluid movement in the container, resulting in a decrease in measurement accuracy. The results indicated that the magnetostrictive liquid level gauge had the potential to be used for water level measurement of SG in nuclear power plant.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":null,"pages":null},"PeriodicalIF":2.7,"publicationDate":"2024-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207227","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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