Pub Date : 2024-09-02DOI: 10.1016/j.net.2024.08.060
S.Y. Zhang, Y.B. Nie, Y.Y. Ding, Q. Zhao, K.Z. Xu, X.Y. Pan, H.T. Chen, Q. Sun, Z. Wei
Integral experiments play a crucial role in advancing nuclear science and technology by providing critical data that validate theoretical models and enhance reactor designs. This study presents a novel approach to accurately determine pulse time distribution in integral experiments conducted with pulsed accelerators. By strategically placed monitors and shields at angles of 0° and 90° relative to the beam direction, neutron flight times from the target are measured, and a response matrix for neutron emission at different times is constructed through simulation. The Maximum Likelihood Expectation Maximization (MLEM) algorithm is employed for pulse time reconstruction, with the gamma ray flight time spectrum from monitors used as the initial spectrum to streamline the computational process. Experimental validation using a standard polyethylene sample and n-p scattering cross-sections confirms the accuracy of the method. Results are compared across multiple nuclear databases such as CENDL-3.2, ENDF/B-VIII.0, JENDL-5.0, and JEFF-3.3 libraries. The developed method significantly enhances the precision of pulse time distribution determination, thereby improving the quality and reliability of experimental data obtained from integral experiments conducted with pulsed accelerators.
{"title":"Accurate pulse time distribution determination using MLEM algorithm in integral experiments","authors":"S.Y. Zhang, Y.B. Nie, Y.Y. Ding, Q. Zhao, K.Z. Xu, X.Y. Pan, H.T. Chen, Q. Sun, Z. Wei","doi":"10.1016/j.net.2024.08.060","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.060","url":null,"abstract":"Integral experiments play a crucial role in advancing nuclear science and technology by providing critical data that validate theoretical models and enhance reactor designs. This study presents a novel approach to accurately determine pulse time distribution in integral experiments conducted with pulsed accelerators. By strategically placed monitors and shields at angles of 0° and 90° relative to the beam direction, neutron flight times from the target are measured, and a response matrix for neutron emission at different times is constructed through simulation. The Maximum Likelihood Expectation Maximization (MLEM) algorithm is employed for pulse time reconstruction, with the gamma ray flight time spectrum from monitors used as the initial spectrum to streamline the computational process. Experimental validation using a standard polyethylene sample and n-p scattering cross-sections confirms the accuracy of the method. Results are compared across multiple nuclear databases such as CENDL-3.2, ENDF/B-VIII.0, JENDL-5.0, and JEFF-3.3 libraries. The developed method significantly enhances the precision of pulse time distribution determination, thereby improving the quality and reliability of experimental data obtained from integral experiments conducted with pulsed accelerators.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207216","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 : 2024-09-02DOI: 10.1016/j.net.2024.08.065
S.F. Quintão Ribeiro, J.J. Rivero
The paper presents a model for operational risk of industrial facilities with applications to nuclear power plants and uranium mining. The traditional models update the risk profile for specific operational conditions setting to one the failure probability of components which are known to be unavailable by failure or maintenance, while constant mean unavailability values from the original Probabilistic Safety Analysis (PSA) are kept for the rest. The proposed methodology considers the time dependency of standby failure probabilities through instantaneous unavailability models, depending on the specific standby time for each component at the given moment, instead of using the same constant values. It also incorporates ageing, testing degradation effects and maintenance effectiveness, not explicitly considered in traditional models, as well as the instantaneous reevaluation of common cause failure probabilities. The results show significant risk underestimations when components specific standby times, ageing, testing degradation effects and maintenance effectiveness are not considered.
{"title":"A model for operational risk of industrial facilities including ageing/test degradation effects and maintenance effectiveness","authors":"S.F. Quintão Ribeiro, J.J. Rivero","doi":"10.1016/j.net.2024.08.065","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.065","url":null,"abstract":"The paper presents a model for operational risk of industrial facilities with applications to nuclear power plants and uranium mining. The traditional models update the risk profile for specific operational conditions setting to one the failure probability of components which are known to be unavailable by failure or maintenance, while constant mean unavailability values from the original Probabilistic Safety Analysis (PSA) are kept for the rest. The proposed methodology considers the time dependency of standby failure probabilities through instantaneous unavailability models, depending on the specific standby time for each component at the given moment, instead of using the same constant values. It also incorporates ageing, testing degradation effects and maintenance effectiveness, not explicitly considered in traditional models, as well as the instantaneous reevaluation of common cause failure probabilities. The results show significant risk underestimations when components specific standby times, ageing, testing degradation effects and maintenance effectiveness are not considered.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"10 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207209","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}
This paper reports the control strategy for the core power in an accelerator drive sub-critical (ADS) system. In an ADS system, the intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. The proposed control strategy is to control the reactor power by adjusting the proton beam power, where the beam power is adjusted by changing either the duty factor or the intensity of the proton beam. As an example, the reactor power control of the China initiative Accelerator Driven System (CiADS) facility has been studied by adjusting the beam power. Firstly, the beam power is set roughly by assigning a new duty factor, where the duty factor is set by changing the beam macro-pulse length and the pulse repetition rate of the proton beam. Both the pulse length and the repetition rate are assigned by a timing system. Secondly, the power is adjusted precisely by changing the beam intensity. To change continuously the beam intensity, an adjustable aperture is used to block the outer particles of the beam line in the accelerator. In order to evaluate the proposed control strategy, a CiADS core model is built based on the multi-node point reactor dynamics model. Three cases, the start of the facility, the decrease of core power and the increase of core power, have been simulated with the model. The simulation results indicate that the control strategy for the core power by changing either the duty factor or the intensity of the proton beam works very well during the operation of the facility.
{"title":"Control strategy for the core power in an accelerator drive sub-critical system","authors":"Xinxin Li, Yuan He, Wenjing Ma, Wenjuan Cui, Zhiyong He, Detai Zhou, Hai Zheng, Feng Yang, Yuhui Guo, Haihua Niu, Kai Yin, Shiwu Dang","doi":"10.1016/j.net.2024.08.066","DOIUrl":"https://doi.org/10.1016/j.net.2024.08.066","url":null,"abstract":"This paper reports the control strategy for the core power in an accelerator drive sub-critical (ADS) system. In an ADS system, the intense external neutron source provided by a proton accelerator coupled to a spallation target is used to drive a sub-critical reactor. The proposed control strategy is to control the reactor power by adjusting the proton beam power, where the beam power is adjusted by changing either the duty factor or the intensity of the proton beam. As an example, the reactor power control of the China initiative Accelerator Driven System (CiADS) facility has been studied by adjusting the beam power. Firstly, the beam power is set roughly by assigning a new duty factor, where the duty factor is set by changing the beam macro-pulse length and the pulse repetition rate of the proton beam. Both the pulse length and the repetition rate are assigned by a timing system. Secondly, the power is adjusted precisely by changing the beam intensity. To change continuously the beam intensity, an adjustable aperture is used to block the outer particles of the beam line in the accelerator. In order to evaluate the proposed control strategy, a CiADS core model is built based on the multi-node point reactor dynamics model. Three cases, the start of the facility, the decrease of core power and the increase of core power, have been simulated with the model. The simulation results indicate that the control strategy for the core power by changing either the duty factor or the intensity of the proton beam works very well during the operation of the facility.","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"7 1","pages":""},"PeriodicalIF":2.7,"publicationDate":"2024-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142207215","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":"10 1","pages":""},"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}
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":"71 1","pages":""},"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}
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":"77 1","pages":""},"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":"14 1","pages":""},"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":"18 1","pages":""},"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":"57 1","pages":""},"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":"6 1","pages":""},"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}
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