2025, 46(1): 1-12.
doi: 10.13832/j.jnpe.2025.01.0001
Abstract:
Amid the global surge in scientific intelligence research and applications, artificial intelligence (AI) technology has been applied in various aspects of reactor neutron analysis to enhance its intelligence, precision and efficiency. This paper provides a comprehensive review of the research progress in the application of AI technologies within reactor neutron analysis, aiming to provide insights for advancing the digitalization of reactors and supporting future developments in this field. We first introduce AI methods' basic classifications and characteristics, and then investigate the four key aspects of neutron analysis: nuclear data evaluation, problem modeling, numerical solution of equations, and transport result application. Also critical technologies are reviewed and summarized for each aspect. Finally, the paper discusses challenges in AI-based neutron analysis, particularly in model, data, and application security. In response to these challenges, the paper proposes research recommendations.
Amid the global surge in scientific intelligence research and applications, artificial intelligence (AI) technology has been applied in various aspects of reactor neutron analysis to enhance its intelligence, precision and efficiency. This paper provides a comprehensive review of the research progress in the application of AI technologies within reactor neutron analysis, aiming to provide insights for advancing the digitalization of reactors and supporting future developments in this field. We first introduce AI methods' basic classifications and characteristics, and then investigate the four key aspects of neutron analysis: nuclear data evaluation, problem modeling, numerical solution of equations, and transport result application. Also critical technologies are reviewed and summarized for each aspect. Finally, the paper discusses challenges in AI-based neutron analysis, particularly in model, data, and application security. In response to these challenges, the paper proposes research recommendations.
2025, 46(1): 13-23.
doi: 10.13832/j.jnpe.2025.01.0013
Abstract:
In response to the evolving demands of advanced small reactor technologies, there is an increasing requirement for three-dimensional whole core transport calculations in reactor physics numerical simulation codes. This paper presents the development of a three-dimensional space-time kinetics neutron transport code, SAAFCGSN, based on the MOOSE platform. The code implements spatial variable discretization using the finite element method, and solves the steady-state and transient neutron transport equations as well as the delayed neutron precursor equations using a residual form approach. The Jacobian-Free Newton-Krylov (JFNK) method is employed to avoid the direct computation of the Jacobian matrix, thereby enhancing computational speed and reducing memory usage. To evaluate the transient computational capability of the code, we validated its reliability using the OECD/NEA C5G7-TD benchmark series and conducted a comparative analysis with high-fidelity deterministic neutron transport codes and Monte Carlo simulations. The study demonstrates that the SAAFCGSN code achieves high computational accuracy, effectively manages the cusping effect of control rods, and obtains detailed neutron flux distributions within the three-dimensional core. It meets the steady-state and transient neutronic calculation requirements for advanced small reactors.
In response to the evolving demands of advanced small reactor technologies, there is an increasing requirement for three-dimensional whole core transport calculations in reactor physics numerical simulation codes. This paper presents the development of a three-dimensional space-time kinetics neutron transport code, SAAFCGSN, based on the MOOSE platform. The code implements spatial variable discretization using the finite element method, and solves the steady-state and transient neutron transport equations as well as the delayed neutron precursor equations using a residual form approach. The Jacobian-Free Newton-Krylov (JFNK) method is employed to avoid the direct computation of the Jacobian matrix, thereby enhancing computational speed and reducing memory usage. To evaluate the transient computational capability of the code, we validated its reliability using the OECD/NEA C5G7-TD benchmark series and conducted a comparative analysis with high-fidelity deterministic neutron transport codes and Monte Carlo simulations. The study demonstrates that the SAAFCGSN code achieves high computational accuracy, effectively manages the cusping effect of control rods, and obtains detailed neutron flux distributions within the three-dimensional core. It meets the steady-state and transient neutronic calculation requirements for advanced small reactors.
2025, 46(1): 47-62.
doi: 10.13832/j.jnpe.2025.01.0047
Abstract:
Neutron radiography shows important application potential in the visualization and measurement of multi-phase flow morphologies in new working medium reactors. This article elaborates on the basic principles of neutron radiography measurement methods, provides a comprehensive review of the research progress of neutron radiography technology in traditional light water reactors, lead-bismuth cooled fast reactors, heat pipe cooled reactors, supercritical water reactors, and sodium-cooled fast reactors. It also outlines the future development directions of neutron radiography technology applied to new working medium reactors and provides the basic methodologies for obtaining high-fidelity neutron images and measuring flow patterns.
Neutron radiography shows important application potential in the visualization and measurement of multi-phase flow morphologies in new working medium reactors. This article elaborates on the basic principles of neutron radiography measurement methods, provides a comprehensive review of the research progress of neutron radiography technology in traditional light water reactors, lead-bismuth cooled fast reactors, heat pipe cooled reactors, supercritical water reactors, and sodium-cooled fast reactors. It also outlines the future development directions of neutron radiography technology applied to new working medium reactors and provides the basic methodologies for obtaining high-fidelity neutron images and measuring flow patterns.
2025, 46(1): 152-159.
doi: 10.13832/j.jnpe.2025.01.0152
Abstract:
The interaction between fuel and matrix caused by asymmetric thermal stress coupling is a key problem in the analysis of solid-core reactor. In this study, a distributed cold source and heat source loading method is developed by combining numerical simulation with experiment, and the thermal stress coupling simulation and experimental study of typical solid-core reactor fuel elements at high temperature are carried out. The research results show that the high temperature strain field measured is close to the numerical simulation result, and there is no risk of failure of the core fuel matrix at 350°C. The numerical prediction method and experimental approach developed in this study for asymmetrically distributed cold and heat source thermal stress coupling can be applied to the analysis of asymmetric thermal stress distribution in solid cores, and there is no risk of failure of the fuel element matrix at low temperature difference.
The interaction between fuel and matrix caused by asymmetric thermal stress coupling is a key problem in the analysis of solid-core reactor. In this study, a distributed cold source and heat source loading method is developed by combining numerical simulation with experiment, and the thermal stress coupling simulation and experimental study of typical solid-core reactor fuel elements at high temperature are carried out. The research results show that the high temperature strain field measured is close to the numerical simulation result, and there is no risk of failure of the core fuel matrix at 350°C. The numerical prediction method and experimental approach developed in this study for asymmetrically distributed cold and heat source thermal stress coupling can be applied to the analysis of asymmetric thermal stress distribution in solid cores, and there is no risk of failure of the fuel element matrix at low temperature difference.
2025, 46(1): 209-215.
doi: 10.13832/j.jnpe.2025.01.0209
Abstract:
The thermal sleeve assembly (referred to as thermal sleeve) is one of the main components of the control rod drive mechanism of the nuclear power plant reactor. Due to the impact of jet fluid on the lower end, the flange and the pressure housing are in contact and wear, which leads to the decrease of bearing capacity and impact resistance and affects the rod dropping function of the control rod. Based on Archard model, this paper puts forward the differential equations of structural wear characteristic time and thermal sleeve wear, decouples structural wear from operating condition parameters and material physical parameters, establishes a static wear analysis model suitable for thermal sleeve, and obtains the variation law of sedimentation with structural wear characteristic time. The results show that for the CRDM structure of HPR 1000, the maximum safety operation life of the thermal sleeve is at an inclination angle with 22.6°. By giving the contrast curve between the remaining operation life and sedimentation height, the evaluation method of thermal sleeve replacement and the treatment scheme of wear defects are provided for the operators of nuclear power plants.
The thermal sleeve assembly (referred to as thermal sleeve) is one of the main components of the control rod drive mechanism of the nuclear power plant reactor. Due to the impact of jet fluid on the lower end, the flange and the pressure housing are in contact and wear, which leads to the decrease of bearing capacity and impact resistance and affects the rod dropping function of the control rod. Based on Archard model, this paper puts forward the differential equations of structural wear characteristic time and thermal sleeve wear, decouples structural wear from operating condition parameters and material physical parameters, establishes a static wear analysis model suitable for thermal sleeve, and obtains the variation law of sedimentation with structural wear characteristic time. The results show that for the CRDM structure of HPR 1000, the maximum safety operation life of the thermal sleeve is at an inclination angle with 22.6°. By giving the contrast curve between the remaining operation life and sedimentation height, the evaluation method of thermal sleeve replacement and the treatment scheme of wear defects are provided for the operators of nuclear power plants.
2025, 46(1): 232-237.
doi: 10.13832/j.jnpe.2025.01.0232
Abstract:
The excessive vibration seriously affects the safe operation of the main pump. To ensure the safe operation of the main pump, root cause analysis technology is adopted to sort out the possible causes of motor vibration of the main pump through fault trees. And the fault mechanism is determined through analysis and searching. Meanwhile the cause was simulated by electromagnetic vibration coupling finite element analysis. Finally, by the adjustment of the position of the stator core and the control of key links, the problem of excessive vibration of the main pump motor was completely solved. After maintenance, the main pump motor has been running continuously for two years, and it has always met the technical requirements for operation, which provides a reference for dealing with the vibration problems of the same type of equipment in the future.
The excessive vibration seriously affects the safe operation of the main pump. To ensure the safe operation of the main pump, root cause analysis technology is adopted to sort out the possible causes of motor vibration of the main pump through fault trees. And the fault mechanism is determined through analysis and searching. Meanwhile the cause was simulated by electromagnetic vibration coupling finite element analysis. Finally, by the adjustment of the position of the stator core and the control of key links, the problem of excessive vibration of the main pump motor was completely solved. After maintenance, the main pump motor has been running continuously for two years, and it has always met the technical requirements for operation, which provides a reference for dealing with the vibration problems of the same type of equipment in the future.
2025, 46(1): 247-253.
doi: 10.13832/j.jnpe.2025.01.0247
Abstract:
The underwater transportation system will be affected by the uncertain nonlinearity of water and other external disturbance when transporting loads. Aiming at the operational control of underwater transportation system, a non-singular terminal sliding mode control method based on radial basis function (RBF) neural network is designed for the underwater transport process of nuclear power plant fuel assembly. Firstly, according to Newton's second law and Morison's equation, the kinetic differential equation of the system is established and its state-space equation is derived. Secondly, a non-singular terminal sliding mode controller is designed, and the unknown nonlinear effect is estimated by RBF neural network and compensated in the controller. The adaptive updating law of network weight is derived by Lyapunov stability theory. The Lyapunov stability theory proves that the proposed control strategy can achieve asymptotic convergence for unknown nonlinear estimation and finite-time convergence for given instruction tracking. Simulations is carried out for the two conditions of upgoing with load and downgoing without load respectively, and the results verify that the controller designed has good performance.
The underwater transportation system will be affected by the uncertain nonlinearity of water and other external disturbance when transporting loads. Aiming at the operational control of underwater transportation system, a non-singular terminal sliding mode control method based on radial basis function (RBF) neural network is designed for the underwater transport process of nuclear power plant fuel assembly. Firstly, according to Newton's second law and Morison's equation, the kinetic differential equation of the system is established and its state-space equation is derived. Secondly, a non-singular terminal sliding mode controller is designed, and the unknown nonlinear effect is estimated by RBF neural network and compensated in the controller. The adaptive updating law of network weight is derived by Lyapunov stability theory. The Lyapunov stability theory proves that the proposed control strategy can achieve asymptotic convergence for unknown nonlinear estimation and finite-time convergence for given instruction tracking. Simulations is carried out for the two conditions of upgoing with load and downgoing without load respectively, and the results verify that the controller designed has good performance.
2025, 46(1): 265-272.
doi: 10.13832/j.jnpe.2025.01.0265
Abstract:
In order to ensure the long-term healthy and stable operation of centrifugal pump, on-line monitoring and fault identification is particularly important. Therefore, this paper presents a fault-identification model based on adaptive noise, with the methods of Complete Ensemble Empirical Mode Decomposition with Adaptive Noise (CEEMDAN), Principal Component Analysis (PCA) and Convolution Neural Network (CNN) combined with Autocorrelation (AC). Firstly, vibration signals are collected and decomposed by CEEMDAN. Then, we discriminate the obtained Intrinsic Mode Functions (IMF) fraction and eliminate the noise fraction to reconstruct the first round of denoised signals. After that, PCA is adopted to remove noise in the denoised signals, and the signals which have been filtered twice are then processed by AC and input to CNN to train the model. Through the experimental verification of a centrifugal pump fault, the results show that compared to traditional methods like double-layer noise reduction with CNN and CEEMD-wavelet denoising-AC-CNN, the model presented in this paper is more resistant to interference and has faster convergence speed. Also, it has the advantages of higher precision and better robustness. At the same order of magnitude, its precision can reach 97.9%.
In order to ensure the long-term healthy and stable operation of centrifugal pump, on-line monitoring and fault identification is particularly important. Therefore, this paper presents a fault-identification model based on adaptive noise, with the methods of Complete Ensemble Empirical Mode Decomposition with Adaptive Noise (CEEMDAN), Principal Component Analysis (PCA) and Convolution Neural Network (CNN) combined with Autocorrelation (AC). Firstly, vibration signals are collected and decomposed by CEEMDAN. Then, we discriminate the obtained Intrinsic Mode Functions (IMF) fraction and eliminate the noise fraction to reconstruct the first round of denoised signals. After that, PCA is adopted to remove noise in the denoised signals, and the signals which have been filtered twice are then processed by AC and input to CNN to train the model. Through the experimental verification of a centrifugal pump fault, the results show that compared to traditional methods like double-layer noise reduction with CNN and CEEMD-wavelet denoising-AC-CNN, the model presented in this paper is more resistant to interference and has faster convergence speed. Also, it has the advantages of higher precision and better robustness. At the same order of magnitude, its precision can reach 97.9%.
2025, 46(1): 289-296.
doi: 10.13832/j.jnpe.2025.01.0289
Abstract:
The nuclear power system of micro reactor with organic Rankine cycle is an effective approach to achieve multiple energy supply in special scenarios. The utilization of organic working fluid lubricated air bearing in organic Rankine cycle turbine generator set can eliminate the complex lubrication oil system, which is an effective measure to improve the compactness of the system. In order to investigate the static characteristics of organic working fluid lubricated air bearing, this paper focuses on the organic working fluid aerostatic thrust bearing, and the influence of inlet pressure, inlet temperature, and rotation speed on loading capacity and mass flow rate is discussed in detail. In addition, the effects of orifice type, axial clearance, and lubricating working fluid are further analyzed. The results show that under different operating conditions, the film pressure distribution characteristics remain basically consistent. The pressure gradually decreases radially from the orifice area to the outlet. Under the conditions with high inlet pressure, high inlet temperature and low rotation speed, the loading capacity can be kept at a high level. Compared to the thrust bearing with small orifice, bearings with slots have higher loading capacity and lower mass flow rate under the same inlet area. As the axial clearance increases, the film pressure significantly decreases, the loading capacity decreases, and the mass flow rate increases. The lubricating working fluid has a certain influence on the static characteristics. In most conditions, bearing lubricated by R134a performs the best static characteristics, followed by cyclopentane bearing, and R245fa bearing presents the relatively worst performance. The results in this paper can provide vital reference for the design optimization of organic working fluid lubricating bearings.
The nuclear power system of micro reactor with organic Rankine cycle is an effective approach to achieve multiple energy supply in special scenarios. The utilization of organic working fluid lubricated air bearing in organic Rankine cycle turbine generator set can eliminate the complex lubrication oil system, which is an effective measure to improve the compactness of the system. In order to investigate the static characteristics of organic working fluid lubricated air bearing, this paper focuses on the organic working fluid aerostatic thrust bearing, and the influence of inlet pressure, inlet temperature, and rotation speed on loading capacity and mass flow rate is discussed in detail. In addition, the effects of orifice type, axial clearance, and lubricating working fluid are further analyzed. The results show that under different operating conditions, the film pressure distribution characteristics remain basically consistent. The pressure gradually decreases radially from the orifice area to the outlet. Under the conditions with high inlet pressure, high inlet temperature and low rotation speed, the loading capacity can be kept at a high level. Compared to the thrust bearing with small orifice, bearings with slots have higher loading capacity and lower mass flow rate under the same inlet area. As the axial clearance increases, the film pressure significantly decreases, the loading capacity decreases, and the mass flow rate increases. The lubricating working fluid has a certain influence on the static characteristics. In most conditions, bearing lubricated by R134a performs the best static characteristics, followed by cyclopentane bearing, and R245fa bearing presents the relatively worst performance. The results in this paper can provide vital reference for the design optimization of organic working fluid lubricating bearings.
2018, 39(5): 1-3.
doi: 10.13832/j.jnpe.2018.05.0001
摘要：
主要介绍了我国在建、在运核电机组的基本状况和最新进展,以及我国在提升核设施安全水平方面的相关措施。在国家能源局印发的《能源技术创新“十三五”规划》要求之下,我国推出一系列先进核能和小型堆的发展计划,开展了“海洋核动力平台示范工程建设”并建立相关标准。最后总结了中国核电目前面临的挑战和未来的展望。
主要介绍了我国在建、在运核电机组的基本状况和最新进展,以及我国在提升核设施安全水平方面的相关措施。在国家能源局印发的《能源技术创新“十三五”规划》要求之下,我国推出一系列先进核能和小型堆的发展计划,开展了“海洋核动力平台示范工程建设”并建立相关标准。最后总结了中国核电目前面临的挑战和未来的展望。
2019, 40(4): 1-8.
摘要：
热管冷却反应堆采用固态反应堆设计理念，通过热管非能动方式导出堆芯热量。本文总结了热管冷却反应堆的概念初创、积极探索、重大突破的发展历程；分析了热管冷却反应堆的技术特点，包括固态属性、固有安全性高、运行特性简单、易于模块化与易扩展和运输特性良好等核心优势；归纳了热管冷却反应堆中热管性能、材料工艺、能量转换等技术现状，并提出热管冷却反应堆进一步发展将面临的材料、制造工艺、运行可维护性等挑战，从而明确了热管冷却反应堆未来的发展趋势，为革新型热管冷却反应堆技术的发展与应用提供良好的方向指引。总体而言，热管冷却反应堆在深空探测与推进、陆基核电源、深海潜航探索等场景中具有广阔的应用前景，有可能成为改变未来核动力格局的颠覆性技术之一。
热管冷却反应堆采用固态反应堆设计理念，通过热管非能动方式导出堆芯热量。本文总结了热管冷却反应堆的概念初创、积极探索、重大突破的发展历程；分析了热管冷却反应堆的技术特点，包括固态属性、固有安全性高、运行特性简单、易于模块化与易扩展和运输特性良好等核心优势；归纳了热管冷却反应堆中热管性能、材料工艺、能量转换等技术现状，并提出热管冷却反应堆进一步发展将面临的材料、制造工艺、运行可维护性等挑战，从而明确了热管冷却反应堆未来的发展趋势，为革新型热管冷却反应堆技术的发展与应用提供良好的方向指引。总体而言，热管冷却反应堆在深空探测与推进、陆基核电源、深海潜航探索等场景中具有广阔的应用前景，有可能成为改变未来核动力格局的颠覆性技术之一。
2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
摘要：
放射性废液得到有效处理是世界各国核工业迅猛发展的前提，其关键技术的现状和发展方向也是我国核工业界关注的焦点。本文介绍了几种放射性废液处理的传统方法及涌现出的新技术，概述了各种方法的原理及优、缺点，同时讨论了放射性废液处理技术今后的研究方向及发展趋势。
放射性废液得到有效处理是世界各国核工业迅猛发展的前提，其关键技术的现状和发展方向也是我国核工业界关注的焦点。本文介绍了几种放射性废液处理的传统方法及涌现出的新技术，概述了各种方法的原理及优、缺点，同时讨论了放射性废液处理技术今后的研究方向及发展趋势。
2019, 40(3): 103-108.
doi: 10.13832/j.jnpe.2019.03.0103
摘要：
以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明：四取中逻辑在可靠性方面优化程度相对较高。
以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明：四取中逻辑在可靠性方面优化程度相对较高。
2019, 40(1): 1-7.
doi: 10.13832/j.jnpe.2019.01.0001
摘要：
“华龙一号”是我国自主设计研发的具有完整知识产权的第三代百万千瓦级压水堆核电技术。本文介绍了“华龙一号”的产生历程，系统论述了“华龙一号”反应堆堆芯与安全设计特点，包括“华龙一号”研发过程中开展的堆芯核设计、热工水力设计、安全设计、设计验证及“华龙一号”持续开展的设计改进与优化等内容，通过采用新的设计理念和设计技术，全面提高了“华龙一号”作为三代核电技术的经济性、灵活性和安全性。
“华龙一号”是我国自主设计研发的具有完整知识产权的第三代百万千瓦级压水堆核电技术。本文介绍了“华龙一号”的产生历程，系统论述了“华龙一号”反应堆堆芯与安全设计特点，包括“华龙一号”研发过程中开展的堆芯核设计、热工水力设计、安全设计、设计验证及“华龙一号”持续开展的设计改进与优化等内容，通过采用新的设计理念和设计技术，全面提高了“华龙一号”作为三代核电技术的经济性、灵活性和安全性。
2020, 41(5): 132-137.
摘要：
为解决核电厂传统监测手段的局限性，提出将核主元分析法（KPCA）引入核电厂设备在线监测领域中，并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性，将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明，KPCA算法可适应核电厂设备监测的要求，能比现有阈值监测手段提供更为早期的故障预警。同时，相比于常规的主元分析法（PCA），KPCA算法能够提取各变量之间的非线性关系，识别出设备不同的运行模式，有效减少误报警。
为解决核电厂传统监测手段的局限性，提出将核主元分析法（KPCA）引入核电厂设备在线监测领域中，并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性，将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明，KPCA算法可适应核电厂设备监测的要求，能比现有阈值监测手段提供更为早期的故障预警。同时，相比于常规的主元分析法（PCA），KPCA算法能够提取各变量之间的非线性关系，识别出设备不同的运行模式，有效减少误报警。
2022, 43(2): 246-253.
doi: 10.13832/j.jnpe.2022.02.0246
摘要：
为了对核电厂主泵的运行过程进行监测和追踪，进而提高主泵的预警能力，提出了基于差分自回归移动平均（ARIMA）和长短期记忆（LSTM）神经网络组合模型的主泵状态预测方法，并用该方法对某核电厂主泵止推轴承温度和可控泄漏流量进行单步和多步预测，以根均方误差（RMSE）为指标对预测精度进行评估。结果表明，所建立的ARIMA和LSTM神经网络组合模型能够对主泵的状态进行准确的预测和追踪，并且组合模型的预测精度要优于ARIMA和LSTM单一模型，尤其在多步预测中，组合模型的优势更加明显。
为了对核电厂主泵的运行过程进行监测和追踪，进而提高主泵的预警能力，提出了基于差分自回归移动平均（ARIMA）和长短期记忆（LSTM）神经网络组合模型的主泵状态预测方法，并用该方法对某核电厂主泵止推轴承温度和可控泄漏流量进行单步和多步预测，以根均方误差（RMSE）为指标对预测精度进行评估。结果表明，所建立的ARIMA和LSTM神经网络组合模型能够对主泵的状态进行准确的预测和追踪，并且组合模型的预测精度要优于ARIMA和LSTM单一模型，尤其在多步预测中，组合模型的优势更加明显。
2019, 40(1): 74-77.
doi: 10.13832/j.jnpe.2019.01.0074
摘要：
介绍了中广核研究院在事故容错燃料（ATF）包壳领域的最新成果，通过预置粉末式脉冲激光熔覆技术，在不同的功率下制备出不同厚度的锆包壳管Cr保护层；通过高温蒸汽氧化增重数据发现，采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能，证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构，采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
介绍了中广核研究院在事故容错燃料（ATF）包壳领域的最新成果，通过预置粉末式脉冲激光熔覆技术，在不同的功率下制备出不同厚度的锆包壳管Cr保护层；通过高温蒸汽氧化增重数据发现，采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能，证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构，采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
2020, 41(3): 137-142.
doi: 10.13832/j.jnpe.2020.03.0137
摘要：
为分析核电厂应急人员在处理严重事故时可能发生的人因失误，通过建立不同应急人员的认知模型及识别相应的行为影响因子，在认知功能的基础上识别出13种人因失误模式：信息来源不足、信息可靠性不佳、过早结束对参数的获取、重要数据处理不正确、缓解措施负面影响评估失误、选择不适用当前情景的策略、延迟决策、遗漏重要信息/警报、延迟发觉、软操作失误、信息反馈失效、设备安装/连接/操作失误、延迟实施，并基于故障树分析得出人因失误模式的主要根原因：交流失效、时间压力、事故发展的不确定性、信息接收延误、监视失误、人-机界面不佳和环境因素。分析结果可用于预测严重事故缓解进程中可能出现的人因失误，为核电厂实施严重事故管理和技术改进，以及保障严重事故工况下核电厂安全提供参考。
为分析核电厂应急人员在处理严重事故时可能发生的人因失误，通过建立不同应急人员的认知模型及识别相应的行为影响因子，在认知功能的基础上识别出13种人因失误模式：信息来源不足、信息可靠性不佳、过早结束对参数的获取、重要数据处理不正确、缓解措施负面影响评估失误、选择不适用当前情景的策略、延迟决策、遗漏重要信息/警报、延迟发觉、软操作失误、信息反馈失效、设备安装/连接/操作失误、延迟实施，并基于故障树分析得出人因失误模式的主要根原因：交流失效、时间压力、事故发展的不确定性、信息接收延误、监视失误、人-机界面不佳和环境因素。分析结果可用于预测严重事故缓解进程中可能出现的人因失误，为核电厂实施严重事故管理和技术改进，以及保障严重事故工况下核电厂安全提供参考。
2020, 41(3): 221-226.
doi: 10.13832/j.jnpe.2020.03.0221
摘要：
为提高已投入运行核动力装置旋转设备的运行数据采集和状态监测能力，需要解决安装传感器和敷设配套线缆困难的问题。本文采用现场可编程门阵列（FPGA）作为主控单元，设计了一种基于Zigbee物联网通信技术的智能无线振动传感器，并给出了其电路构成、工作原理，以及嵌入式控制软件的工作流程。通过对此传感器进行性能测试，结果表明该传感器功耗低，实现了对振动信号的连续采集、智能分析与上传。该无线传感器安装简单，无需敷设供电和信号线缆，可应用于构建核动力装置旋转设备的状态监测系统。
为提高已投入运行核动力装置旋转设备的运行数据采集和状态监测能力，需要解决安装传感器和敷设配套线缆困难的问题。本文采用现场可编程门阵列（FPGA）作为主控单元，设计了一种基于Zigbee物联网通信技术的智能无线振动传感器，并给出了其电路构成、工作原理，以及嵌入式控制软件的工作流程。通过对此传感器进行性能测试，结果表明该传感器功耗低，实现了对振动信号的连续采集、智能分析与上传。该无线传感器安装简单，无需敷设供电和信号线缆，可应用于构建核动力装置旋转设备的状态监测系统。
2022, 43(1): 1-6.
doi: 10.13832/j.jnpe.2022.01.0001
Abstract:
2020, 41(4): 1-7.
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2023, 44(1): 1-8.
doi: 10.13832/j.jnpe.2023.01.0001
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2021, 42(4): 1-5.
doi: 10.13832/j.jnpe.2021.04.0001
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2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
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2022, 43(4): 1-4.
doi: 10.13832/j.jnpe.2022.04.0001
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2019, 40(4): 1-8.
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2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
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2018, 39(3): 6-9.
doi: 10.13832/j.jnpe.2018.03.0006
Abstract:
2023, 44(2): 1-8.
doi: 10.13832/j.jnpe.2023.02.0001
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