cometadata/datacite-titles-descriptions-related-identifiers huggingface dataset:
DOI: 10.57760/sciencedb.04208
title: Updated design of water-cooled breeder blanket for CFETR
title: Updated design of water-cooled breeder blanket for CFETR

description: The water-cooled ceramic breeder (WCCB) blanket is one candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), and it is being developed at the Institute of Plasma Physics of the Chinese Academy of Sciences. This paper presents an updated WCCB blanket concept design that is expected to cover the CFETR operation modes of 200 MW, 500 MW, 1 GW, and 1.5 GW fusion power and achieve tritium self-sufficiency with the latest core parameters (major radius R = 7.2 m; minor radius a = 2.2 m) and mission. The feasibility of the new blanket design is evaluated considering neutronics and thermal-hydraulics aspects.
description: The water-cooled ceramic breeder (WCCB) blanket is one candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), and it is being developed at the Institute of Plasma Physics of the Chinese Academy of Sciences. This paper presents an updated WCCB blanket concept design that is expected to cover the CFETR operation modes of 200 MW, 500 MW, 1 GW, and 1.5 GW fusion power and achieve tritium self-sufficiency with the latest core parameters (major radius R = 7.2 m; minor radius a = 2.2 m) and mission. The feasibility of the new blanket design is evaluated considering neutronics and thermal-hydraulics aspects.
DOI: 10.57760/sciencedb.04211
title: Application of dynamic fault tree in reliability assessment of reactor protection system
title: 动态故障树在反应堆保护系统可靠性评估中的实践应用

description: [Background] Reactor protection system (RPS) has complex structure, numerous equipment, and continuous failure behaviors. The traditional static fault tree cannot accurately simulate the sequential failure behaviors of the RPS. [Purpose] This study aims to obtain the failure probability of related system of RPS and the sensitivity of each component by simulating the sequential failure behaviors of RPS, and provide suggestions for the optimal design and optimization of RPS in the future. [Methods] First of all, the dynamic fault tree (DFT) modeling tools was employed to model the reliability of the RPS, a dynamic binary tree method was applied to quantitative analysis of the DFT model. Then, the Latin hypercube sampling method was adopted to quantitatively analyze the uncertainty of the RPS failure probability, and compared with the traditional static fault tree method under the same conditions. [Results] Simulation results show that the failure probability with 0.95 confidence interval of RPS is 0.015 306~0.015 721. The two most sensitive components are the undervoltage drive card and the undervoltage relay. [Conclusions] DFT can accurately simulate the reliability of RPS, and provide more accurate results, which is beneficial for optimization design in the future.
description: [Background] Reactor protection system (RPS) has complex structure, numerous equipment, and continuous failure behaviors. The traditional static fault tree cannot accurately simulate the sequential failure behaviors of the RPS. [Purpose] This study aims to obtain the failure probability of related system of RPS and the sensitivity of each component by simulating the sequential failure behaviors of RPS, and provide suggestions for the optimal design and optimization of RPS in the future. [Methods] First of all, the dynamic fault tree (DFT) modeling tools was employed to model the reliability of the RPS, a dynamic binary tree method was applied to quantitative analysis of the DFT model. Then, the Latin hypercube sampling method was adopted to quantitatively analyze the uncertainty of the RPS failure probability, and compared with the traditional static fault tree method under the same conditions. [Results] Simulation results show that the failure probability with 0.95 confidence interval of RPS is 0.015 306~0.015 721. The two most sensitive components are the undervoltage drive card and the undervoltage relay. [Conclusions] DFT can accurately simulate the reliability of RPS, and provide more accurate results, which is beneficial for optimization design in the future.
DOI: 10.57760/sciencedb.04212
title: A data-driven adaptive fault diagnosis methodology for nuclear power systems based on NSGAII-CNN
title: A data-driven adaptive fault diagnosis methodology for nuclear power systems based on NSGAII-CNN

description: With the development of digital information technology, nuclear energy systems are developing in the direction of intelligence and unmanned, which requires a higher demand for its safety, such as autonomous fault diagnosis. At present, the network structure model used in fault diagnosis usually needs professional design, which is time-consuming and labor-intensive, and the efficiency is low. To solve these problems, this paper proposes a data-driven adaptive fault diagnosis approach NSGAII-CNN. Firstly, the time-series data are mapped into two-dimensional images by Markov Transition Field, which preserves the time characteristics of the data and improves the fault diagnosis accuracy. Then, the NSGAII-CNN algorithm is proposed to realize the self-adaptive search of the network structure, which improves the construction speed of the fault diagnosis network structure model, thereby improving the diagnosis accuracy and efficiency. Finally, compared with the current three classical CNN architecture models designed by professionals, the methodology proposed in this paper has significant advantages in fault diagnosis and model structure construction. The proposed diagnosis method will provide operators with useful information and enhance the nuclear energy systems’ self-diagnostic capabilities.
description: With the development of digital information technology, nuclear energy systems are developing in the direction of intelligence and unmanned, which requires a higher demand for its safety, such as autonomous fault diagnosis. At present, the network structure model used in fault diagnosis usually needs professional design, which is time-consuming and labor-intensive, and the efficiency is low. To solve these problems, this paper proposes a data-driven adaptive fault diagnosis approach NSGAII-CNN. Firstly, the time-series data are mapped into two-dimensional images by Markov Transition Field, which preserves the time characteristics of the data and improves the fault diagnosis accuracy. Then, the NSGAII-CNN algorithm is proposed to realize the self-adaptive search of the network structure, which improves the construction speed of the fault diagnosis network structure model, thereby improving the diagnosis accuracy and efficiency. Finally, compared with the current three classical CNN architecture models designed by professionals, the methodology proposed in this paper has significant advantages in fault diagnosis and model structure construction. The proposed diagnosis method will provide operators with useful information and enhance the nuclear energy systems’ self-diagnostic capabilities.
DOI: 10.57760/sciencedb.04224
title: data for ‘A Novel Microfluidic Device Integrated with Chitosan‐Modified Capillaries for Rapid ZIKV Detection’
title: data for ‘A Novel Microfluidic Device Integrated with Chitosan‐Modified Capillaries for Rapid ZIKV Detection’

description: These data are original data of the paper 'A Novel Microfluidic Device Integrated with Chitosan‐Modified Capillaries for Rapid ZIKV Detection'.
description: These data are original data of the paper 'A Novel Microfluidic Device Integrated with Chitosan‐Modified Capillaries for Rapid ZIKV Detection'.
DOI: 10.57760/sciencedb.04215
title: Development of cosVMPT and Application of Creating 3D Neutronics Model for 360-Degree CFETR
title: Development of cosVMPT and Application of Creating 3D Neutronics Model for 360-Degree CFETR

description: In order to conduct nuclear analyses for Chinese Fusion Engineering Testing Reactor (CFETR) according to the latest design, it is necessary to establish a 3D neutronics model based on the engineering CAD model. The 360-degree CFETR engineering CAD model is an extremely complex and huge model, with a large number of spline surfaces, flow channels, and overlaps, which puts forward higher requirements for the modeling and the conversion code. In this paper, the detailed modeling and conversion process is introduced, and a detailed 3D neutronics model for 360-degree CFETR is created by cosVMPT (Visual Modeling Platform for Particle Transport). Furthermore, considering the update and maintenance of the CAD model in the future, a new function of ‘‘User-Defined-Void’’ is developed in cosVMPT. The converted geometry is validated by comparing results of volume calculations by MCNP and CAD system. The result shows that the generated three-dimensional neutronics model is in good agreement with the original CAD model. This study demonstrates the capability of cosVMPT to be applied to the nuclear device with large scale complex geometry.
description: In order to conduct nuclear analyses for Chinese Fusion Engineering Testing Reactor (CFETR) according to the latest design, it is necessary to establish a 3D neutronics model based on the engineering CAD model. The 360-degree CFETR engineering CAD model is an extremely complex and huge model, with a large number of spline surfaces, flow channels, and overlaps, which puts forward higher requirements for the modeling and the conversion code. In this paper, the detailed modeling and conversion process is introduced, and a detailed 3D neutronics model for 360-degree CFETR is created by cosVMPT (Visual Modeling Platform for Particle Transport). Furthermore, considering the update and maintenance of the CAD model in the future, a new function of ‘‘User-Defined-Void’’ is developed in cosVMPT. The converted geometry is validated by comparing results of volume calculations by MCNP and CAD system. The result shows that the generated three-dimensional neutronics model is in good agreement with the original CAD model. This study demonstrates the capability of cosVMPT to be applied to the nuclear device with large scale complex geometry.
DOI: 10.57760/sciencedb.04219
title: Development of an assistant program for CAD-to-cosRMC modelling
title: Development of an assistant program for CAD-to-cosRMC modelling

description: cosRMC is a three dimensional neutron photon transport Monte Carlo code based on the stochastic statistical method of neutral particle transport. At present, the cosRMC geometry model is imported by writing the input file manually, which leads to time-consuming, inefficient and error-prone problems. This will limit application of cosRMC in fusion neutronics calculations because of the complex geometry in fusion devices. Therefore, in order to improve the modeling efficiency and meet the requirements of calculation and analysis of advanced nuclear facilities, a visual modeling platform, cosVMPT—Visual Modeling Platform for Particle Transport, is developed based on Open Cascade (OCC) as CAD geometry engine and FreeCAD framework as graphical user interface. This platform can realize the complete process from modeling to calculation, then to visualization. In this article, the conversion algorithm from CAD models to cosRMC geometry models is introduced in detail. The reliability of the conversion algorithm is preliminarily verified with some representative models by comparing volumes change before and after conversion. The result shows that the geometries after conversion are in good consistency with the original geometries.
description: cosRMC is a three dimensional neutron photon transport Monte Carlo code based on the stochastic statistical method of neutral particle transport. At present, the cosRMC geometry model is imported by writing the input file manually, which leads to time-consuming, inefficient and error-prone problems. This will limit application of cosRMC in fusion neutronics calculations because of the complex geometry in fusion devices. Therefore, in order to improve the modeling efficiency and meet the requirements of calculation and analysis of advanced nuclear facilities, a visual modeling platform, cosVMPT—Visual Modeling Platform for Particle Transport, is developed based on Open Cascade (OCC) as CAD geometry engine and FreeCAD framework as graphical user interface. This platform can realize the complete process from modeling to calculation, then to visualization. In this article, the conversion algorithm from CAD models to cosRMC geometry models is introduced in detail. The reliability of the conversion algorithm is preliminarily verified with some representative models by comparing volumes change before and after conversion. The result shows that the geometries after conversion are in good consistency with the original geometries.
DOI: 10.57760/sciencedb.04221
title: Benchmark of cosRMC with CFETR fusion neutronics model
title: Benchmark of cosRMC with CFETR fusion neutronics model

description: cosRMC is a three dimensional Monte Carlo code for coupled neutron and photon transport simulation, it is jointly developed for fission reactor by State Power Investment Corporation Research Institute of China, Tsinghua University and North China Electric Power University. In order to evaluate its feasibility and accuracy in fusion neutronics analysis, a preliminary benchmark is conducted on cosRMC using CFETR fusion neutronics model with the WCCB blanket module. The tritium breeding ratio (TBR), nuclear heating (NH) and neutron wall loading (NWL) are calculated by cosRMC, and the results are compared with those calculated by MCNP with the same nuclear data library FENDL-2.1. Preliminary comparison results show that a very good consistency is obtained between the two Monte Carlo codes: the relative differences of calculated results are mostly within 1% and the statistical errors are within the statistical uncertainty of the MC method.
description: cosRMC is a three dimensional Monte Carlo code for coupled neutron and photon transport simulation, it is jointly developed for fission reactor by State Power Investment Corporation Research Institute of China, Tsinghua University and North China Electric Power University. In order to evaluate its feasibility and accuracy in fusion neutronics analysis, a preliminary benchmark is conducted on cosRMC using CFETR fusion neutronics model with the WCCB blanket module. The tritium breeding ratio (TBR), nuclear heating (NH) and neutron wall loading (NWL) are calculated by cosRMC, and the results are compared with those calculated by MCNP with the same nuclear data library FENDL-2.1. Preliminary comparison results show that a very good consistency is obtained between the two Monte Carlo codes: the relative differences of calculated results are mostly within 1% and the statistical errors are within the statistical uncertainty of the MC method.
DOI: 10.57760/sciencedb.04228
title: Operation and shutdown dose rate analysis of CFETR ECRH system
title: Operation and shutdown dose rate analysis of CFETR ECRH system

description: Electron Cyclotron Resonance Heating (ECRH) is one of auxiliary heating systems to be installed in CFETR. For heating purpose of ECRH, a penetration space has been designed to transmit the wave to plasma. This will significantly increase the neutrons streaming through the wave channel and result in heavy irradiation of ECRH port structures. For nuclear safety analysis, the irradiation dose during operation of CFETR should be limited according to the regulation of radiological zone. In case of failure, the ECRH is planned to be maintained by workers after shutdown and waiting time of one month until the dose rate decreases to low level of 100 μSv/h. The shielding capabilities of ECRH port has been evaluated in view of dose effect both during operation and after shutdown of CFETR. For that purpose, a 3-D neutronics model has been built to launch the nuclear analysis. The operation dose rate was directly transferred from the neutron and photon flux by using of ICRP-74 flux to dose conversion factors. The shutdown dose rate was obtained with NASCA code system by coupling the Monte Carlo particle transport and FISPACT activation calculation. The obtained operation and shutdown dose data can be provided for the shielding optimization of ECRH wave system.
description: Electron Cyclotron Resonance Heating (ECRH) is one of auxiliary heating systems to be installed in CFETR. For heating purpose of ECRH, a penetration space has been designed to transmit the wave to plasma. This will significantly increase the neutrons streaming through the wave channel and result in heavy irradiation of ECRH port structures. For nuclear safety analysis, the irradiation dose during operation of CFETR should be limited according to the regulation of radiological zone. In case of failure, the ECRH is planned to be maintained by workers after shutdown and waiting time of one month until the dose rate decreases to low level of 100 μSv/h. The shielding capabilities of ECRH port has been evaluated in view of dose effect both during operation and after shutdown of CFETR. For that purpose, a 3-D neutronics model has been built to launch the nuclear analysis. The operation dose rate was directly transferred from the neutron and photon flux by using of ICRP-74 flux to dose conversion factors. The shutdown dose rate was obtained with NASCA code system by coupling the Monte Carlo particle transport and FISPACT activation calculation. The obtained operation and shutdown dose data can be provided for the shielding optimization of ECRH wave system.
DOI: 10.57760/sciencedb.04230
title: Neutronics analyses of water-cooled ceramic breeder blanket for the updated CFETR model
title: Neutronics analyses of water-cooled ceramic breeder blanket for the updated CFETR model

description: Based on a three-dimensional water-cooled ceramic breeder (WCCB) blanket model for the China Fusion Engineering Test Reactor (CFETR), neutronics analyses are performed by utilizing the MCNP5 code and the IAEA Fusion Evaluated Nuclear Data Library FENDL2.1. The results include neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heating power and radiation damage under 200MW, 500MW, 1.0GW, 1.5GW fusion power. The results show that the neutronics analyses of WCCB meet the design requirements of CFETR.
description: Based on a three-dimensional water-cooled ceramic breeder (WCCB) blanket model for the China Fusion Engineering Test Reactor (CFETR), neutronics analyses are performed by utilizing the MCNP5 code and the IAEA Fusion Evaluated Nuclear Data Library FENDL2.1. The results include neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heating power and radiation damage under 200MW, 500MW, 1.0GW, 1.5GW fusion power. The results show that the neutronics analyses of WCCB meet the design requirements of CFETR.
DOI: 10.57760/sciencedb.04234
title: Electromagnetic force evaluation on the water cooled ceramic breeder blanket for CFETR
title: Electromagnetic force evaluation on the water cooled ceramic breeder blanket for CFETR

description: The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). To investigate the electromagnetic (EM) force (magnetization force and Lorentz force) on the WCCB blanket, EM analyses were carried out using ANSYS Multiphysics code. At fifirst, fifinite element (FE) static magnetic analysis under normal operation condition was accomplished, adopting the magnetic scalar potential (MSP) method. The magnetization force can be obtained as a result. Then transient magnetic analysis under plasma disruption event was performed using the magnetic vector potential (MVP) method. The eddy current on blankets was subsequently analyzed. Finally, the Lorentz force applied on blankets can be calculated with cross-product of the eddy current fifield from transient magnetic analysis and the magnetic fifield from the static magnetic analysis. The EM forces will be used as mechanical loads for the future structural analysis of blanket under integrated load.
description: The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). To investigate the electromagnetic (EM) force (magnetization force and Lorentz force) on the WCCB blanket, EM analyses were carried out using ANSYS Multiphysics code. At fifirst, fifinite element (FE) static magnetic analysis under normal operation condition was accomplished, adopting the magnetic scalar potential (MSP) method. The magnetization force can be obtained as a result. Then transient magnetic analysis under plasma disruption event was performed using the magnetic vector potential (MVP) method. The eddy current on blankets was subsequently analyzed. Finally, the Lorentz force applied on blankets can be calculated with cross-product of the eddy current fifield from transient magnetic analysis and the magnetic fifield from the static magnetic analysis. The EM forces will be used as mechanical loads for the future structural analysis of blanket under integrated load.
DOI: 10.57760/sciencedb.04236
title: Analysis and comparison of low activation characteristics of nuclear structural material SIMP steel
title: Analysis and comparison of low activation characteristics of nuclear structural material SIMP steel

description: The activation properties of SIMP steel have been simulated within China Fusion Engineering Test Reactor (CFETR) water cooled ceramic breeder (WCCB) blanket. Supposing that the water cooled blanket uses SIMP steel, the simulation is conducted by coupling of the MCNP (Monte Carlo N-Particle Transport Code) and FISPACT activation code. The simulation results of specific activity, decay heat, contact dose rate and radiation damage of SIMP steel have been fully compared with other RAFM steel such as EUROFER-97, F82H and so on. The comparisons show that the SIMP has good activation performance corresponding to low activation performance of RAFM. It is conclusion that the SIMP can be chosen as one candidate structural material used for future fusion reactor blanket.
description: The activation properties of SIMP steel have been simulated within China Fusion Engineering Test Reactor (CFETR) water cooled ceramic breeder (WCCB) blanket. Supposing that the water cooled blanket uses SIMP steel, the simulation is conducted by coupling of the MCNP (Monte Carlo N-Particle Transport Code) and FISPACT activation code. The simulation results of specific activity, decay heat, contact dose rate and radiation damage of SIMP steel have been fully compared with other RAFM steel such as EUROFER-97, F82H and so on. The comparisons show that the SIMP has good activation performance corresponding to low activation performance of RAFM. It is conclusion that the SIMP can be chosen as one candidate structural material used for future fusion reactor blanket.
DOI: 10.57760/sciencedb.04237
title: A Scale-invariant WCNS Schemes 2022
title: 尺度不变的WCNS格式2022

description: Writing code with FotranPost-processing with TecplotPaper preparation and layout using Word
description: Writing code with FotranPost-processing with TecplotPaper preparation and layout using Word