2025, 46(3): 1-17.
doi: 10.13832/j.jnpe.2025.02.0062
Abstract:
Monte Carlo (MC) particle transport methodology incorporates stochastic principles derived from probability theory and mathematical statistics to establish computational frameworks. This approach facilitates the numerical resolution of complex particle transport phenomena in nuclear systems. Over the course of seven decades of development, MC particle transport theory and algorithms have reached a high level of technical maturity. This has resulted in the development of several specialized software packages, which are widely applied in fields such as nuclear radiation shielding, reactor core criticality safety analysis, nuclear detection, and radiation medicine. This study commences by establishing the theoretical framework underlying MC particle transport methodologies. Through rigorous mathematical derivation, we present the neutron flux density formulation developed via MC simulations for addressing integral-form neutron transport equations, coupled with analytical frameworks for determining associated response parameters. It also outlines the classification of deterministic approaches for solving transport equations. The study reviews the historical development and computational application of MC particle transport methods, while summarizing significant software developed domestically and internationally. Furthermore, it examines recent advancements in utilizing graphics processing unit (GPU) technology to develop MC particle transport software, highlighting current research directions and progress in this field. This paper provides a comprehensive review of recent advancements in MC particle transport methodologies and associated software, with a specific focus on key features and capabilities of the independently developed J Monte Carlo transport (JMCT) software.
Monte Carlo (MC) particle transport methodology incorporates stochastic principles derived from probability theory and mathematical statistics to establish computational frameworks. This approach facilitates the numerical resolution of complex particle transport phenomena in nuclear systems. Over the course of seven decades of development, MC particle transport theory and algorithms have reached a high level of technical maturity. This has resulted in the development of several specialized software packages, which are widely applied in fields such as nuclear radiation shielding, reactor core criticality safety analysis, nuclear detection, and radiation medicine. This study commences by establishing the theoretical framework underlying MC particle transport methodologies. Through rigorous mathematical derivation, we present the neutron flux density formulation developed via MC simulations for addressing integral-form neutron transport equations, coupled with analytical frameworks for determining associated response parameters. It also outlines the classification of deterministic approaches for solving transport equations. The study reviews the historical development and computational application of MC particle transport methods, while summarizing significant software developed domestically and internationally. Furthermore, it examines recent advancements in utilizing graphics processing unit (GPU) technology to develop MC particle transport software, highlighting current research directions and progress in this field. This paper provides a comprehensive review of recent advancements in MC particle transport methodologies and associated software, with a specific focus on key features and capabilities of the independently developed J Monte Carlo transport (JMCT) software.
2025, 46(3): 18-23.
doi: 10.13832/j.jnpe.2024.10.0029
Abstract:
To support the research and development of heat pipe reactors, this study designed and constructed a high-temperature compressed air cooling experimental platform to investigate the startup characteristics of high length-diameter arterial sodium heat pipes. The analysis results demonstrate the following: ① In the initial stage of the heat pipe startup process, high-temperature compressed air elevates the temperature of the condensation section, which facilitates the formation of a continuous flow of sodium vapor within the heat pipe, thereby accelerating the cold-state startup speed of the heat pipe; ② During the startup process, preheating of the condensation section enhances the temperature of the sodium vapor, effectively preventing the occurrence of sonic limit phenomenon, and consequently, increases the probability of successful heat pipe startup. The results in this paper provide data and theoretical support for the optimization of the cold start-up mode of the arterial sodium heat pipe with large length-diameter ratio.
To support the research and development of heat pipe reactors, this study designed and constructed a high-temperature compressed air cooling experimental platform to investigate the startup characteristics of high length-diameter arterial sodium heat pipes. The analysis results demonstrate the following: ① In the initial stage of the heat pipe startup process, high-temperature compressed air elevates the temperature of the condensation section, which facilitates the formation of a continuous flow of sodium vapor within the heat pipe, thereby accelerating the cold-state startup speed of the heat pipe; ② During the startup process, preheating of the condensation section enhances the temperature of the sodium vapor, effectively preventing the occurrence of sonic limit phenomenon, and consequently, increases the probability of successful heat pipe startup. The results in this paper provide data and theoretical support for the optimization of the cold start-up mode of the arterial sodium heat pipe with large length-diameter ratio.
2025, 46(3): 137-146.
doi: 10.13832/j.jnpe.2024.060022
Abstract:
To solve the problem that the heat exchange efficiency of alloy 690 heat transfer tube is difficult to reach the design value, this paper adopts the method of combining simulation and experiment, employs the coupled electromagnetic treatment to study the thermal conductivity and mechanical properties of alloy 690 heat transfer tube by applying electric and magnetic fields with different parameters. The results show that when the applied electromagnetic field parameters are 1.5 V-1.5 T, the thermal conductivity of alloy 690 heat transfer tube is increased by 19.6%, and the tensile strength and Vickers hardness are also increased by 6.8% and 4.3%, respectively. The thermal stress calculated by simulation is an order of magnitude larger than the modified Peierls stress, which shows that the coupled electromagnetic treatment can effectively drive the internal dislocation movement of alloy 690. EDS results showed that the coupled electromagnetic energy field could promote the precipitation of intergranular carbides (M23C6), thereby improving the thermal conductivity of alloy 690 heat transfer tubes. In this paper, the feasibility of the coupled electromagnetic treatment to improve the thermal conductivity of alloy 690 heat transfer tube is fully verified, and the heat exchange efficiency of alloy 690 heat transfer tube can be effectively improved.
To solve the problem that the heat exchange efficiency of alloy 690 heat transfer tube is difficult to reach the design value, this paper adopts the method of combining simulation and experiment, employs the coupled electromagnetic treatment to study the thermal conductivity and mechanical properties of alloy 690 heat transfer tube by applying electric and magnetic fields with different parameters. The results show that when the applied electromagnetic field parameters are 1.5 V-1.5 T, the thermal conductivity of alloy 690 heat transfer tube is increased by 19.6%, and the tensile strength and Vickers hardness are also increased by 6.8% and 4.3%, respectively. The thermal stress calculated by simulation is an order of magnitude larger than the modified Peierls stress, which shows that the coupled electromagnetic treatment can effectively drive the internal dislocation movement of alloy 690. EDS results showed that the coupled electromagnetic energy field could promote the precipitation of intergranular carbides (M23C6), thereby improving the thermal conductivity of alloy 690 heat transfer tubes. In this paper, the feasibility of the coupled electromagnetic treatment to improve the thermal conductivity of alloy 690 heat transfer tube is fully verified, and the heat exchange efficiency of alloy 690 heat transfer tube can be effectively improved.
2025, 46(3): 166-172.
doi: 10.13832/j.jnpe.2024.070035
Abstract:
To improve the automation level of PWR nuclear power plant units during startup, reduce the work intensity of reactor operators, shorten the start-up time, and improve the correctness and standardization of the unit start-up, this study puts forward a control technology suitable for automatic startup of nuclear power plant, which is based on the characteristics of typical PWR nuclear power plant unit system, operation management process and control requirements of automatic startup. Based on the analysis of the applicable control range, operation breakpoints, sequence control and analog control of the automatic start-up control system of PWR, an architecture for the automatic start-up control system of nuclear power plant is established, including the architecture design, the functions and design contents of each level and the interactive interface design between levels. At the same time, a typical PWR nuclear power plant automatic start-up simulation platform is established, and the automatic start-up control system is designed by taking the start-up process of nuclear power plant operation mode Ⅲ as an example, and the proposed technical scheme is simulated and verified. The simulation results show that the automatic start-up control system design can realize the automatic start-up of nuclear power plant mode Ⅲ, reducing the operation steps and workload of operators. The designed automatic start-up control system architecture in this study can provide a reference for the application of automatic start-up control system for nuclear power plant, and is of great significance for improving the automation level of start-up process of nuclear power plant units.
To improve the automation level of PWR nuclear power plant units during startup, reduce the work intensity of reactor operators, shorten the start-up time, and improve the correctness and standardization of the unit start-up, this study puts forward a control technology suitable for automatic startup of nuclear power plant, which is based on the characteristics of typical PWR nuclear power plant unit system, operation management process and control requirements of automatic startup. Based on the analysis of the applicable control range, operation breakpoints, sequence control and analog control of the automatic start-up control system of PWR, an architecture for the automatic start-up control system of nuclear power plant is established, including the architecture design, the functions and design contents of each level and the interactive interface design between levels. At the same time, a typical PWR nuclear power plant automatic start-up simulation platform is established, and the automatic start-up control system is designed by taking the start-up process of nuclear power plant operation mode Ⅲ as an example, and the proposed technical scheme is simulated and verified. The simulation results show that the automatic start-up control system design can realize the automatic start-up of nuclear power plant mode Ⅲ, reducing the operation steps and workload of operators. The designed automatic start-up control system architecture in this study can provide a reference for the application of automatic start-up control system for nuclear power plant, and is of great significance for improving the automation level of start-up process of nuclear power plant units.
2025, 46(3): 186-194.
doi: 10.13832/j.jnpe.2024.070033
Abstract:
To investigate the thermal-hydraulic performance of lead-bismuth centrifugal pump in a closed-loop system transporting 400℃ liquid Lead-Bismuth eutectic (LBE), a joint simplified modeling approach that integrates the lead-bismuth circulation tank, inlet/outlet pipelines, and centrifugal pump was employed. Utilizing the shear stress transport (SST) k-ω turbulence model, flow characteristics within the pump under three distinct flow rate conditions were systematically analyzed. The study revealed that vortices of varying intensities in the impeller flow channels were associated with fluid force imbalances, with Coriolis forces maintaining dominant influence throughout the LBE transport process. Local entropy production rate (EPR) was primarily concentrated at the leading edge of impeller blades and rotor-stator interface regions, exhibiting a decreasing trend with increasing flow rates. Pressure signal frequencies in the impeller and guide vane channels demonstrated periodic alternations between 93.33 Hz and 116.67 Hz, while wavelet signal intensity became more pronounced near the rotor-stator interface. These findings provide important references for optimizing design and performance evaluation of centrifugal pumps in lead-bismuth reactor systems.
To investigate the thermal-hydraulic performance of lead-bismuth centrifugal pump in a closed-loop system transporting 400℃ liquid Lead-Bismuth eutectic (LBE), a joint simplified modeling approach that integrates the lead-bismuth circulation tank, inlet/outlet pipelines, and centrifugal pump was employed. Utilizing the shear stress transport (SST) k-ω turbulence model, flow characteristics within the pump under three distinct flow rate conditions were systematically analyzed. The study revealed that vortices of varying intensities in the impeller flow channels were associated with fluid force imbalances, with Coriolis forces maintaining dominant influence throughout the LBE transport process. Local entropy production rate (EPR) was primarily concentrated at the leading edge of impeller blades and rotor-stator interface regions, exhibiting a decreasing trend with increasing flow rates. Pressure signal frequencies in the impeller and guide vane channels demonstrated periodic alternations between 93.33 Hz and 116.67 Hz, while wavelet signal intensity became more pronounced near the rotor-stator interface. These findings provide important references for optimizing design and performance evaluation of centrifugal pumps in lead-bismuth reactor systems.
2025, 46(3): 224-228.
doi: 10.13832/j.jnpe.2024.070025
Abstract:
By reviewing the design requirements of periodic tests in HAF102 and GB/T5204, combing the characteristics of HPR1000 reactor protection system (RPR) in Units 1 and 2 of Zhangzhou Nuclear Power Plant, Fujian Province, and using the idea of full link coverage and test segment overlapping, a complete set of RPR periodic test design based on NASPIC for the HPR1000 is proposed. Compared with the test schemes of other nuclear power units in China, this scheme adopts automation and human-friendly design for optimization and improvement on the basis of meeting the functional requirements of RPR system, achieving automated execution of tests and reducing the risks brought by human factors. This can provide a reference for the periodic testing schemes of RPR system in subsequent projects.
By reviewing the design requirements of periodic tests in HAF102 and GB/T5204, combing the characteristics of HPR1000 reactor protection system (RPR) in Units 1 and 2 of Zhangzhou Nuclear Power Plant, Fujian Province, and using the idea of full link coverage and test segment overlapping, a complete set of RPR periodic test design based on NASPIC for the HPR1000 is proposed. Compared with the test schemes of other nuclear power units in China, this scheme adopts automation and human-friendly design for optimization and improvement on the basis of meeting the functional requirements of RPR system, achieving automated execution of tests and reducing the risks brought by human factors. This can provide a reference for the periodic testing schemes of RPR system in subsequent projects.
2025, 46(3): 271-281.
doi: 10.13832/j.jnpe.2024.070006
Abstract:
To solve the problem that traditional PWR nuclear power units cannot meet the need to participate in grid peak regulation in the future, an Solar-Nuclear-Storage Hybrid System that couples solar energy and nuclear energy was put forward. A system model was built with thermal system simulation software EBSILON, where exergy analysis of the system under different operation strategies was carried out to study the thermodynamic performance of the system under design conditions. The exergy analysis and research under different operation strategies show that the three equipment with the highest exergy losses in the system are steam generator, solar field and steam turbine high pressure cylinder first stage, and the exergy loss of the three equipment in total is close to 50% of the total exergy loss. At the same time, the exergic efficiency of solar field is mainly affected by the change of direct normal irradiance. The exergic efficiency of electric heater is basically unchanged with the maximum change of 2% under different operation strategies.
To solve the problem that traditional PWR nuclear power units cannot meet the need to participate in grid peak regulation in the future, an Solar-Nuclear-Storage Hybrid System that couples solar energy and nuclear energy was put forward. A system model was built with thermal system simulation software EBSILON, where exergy analysis of the system under different operation strategies was carried out to study the thermodynamic performance of the system under design conditions. The exergy analysis and research under different operation strategies show that the three equipment with the highest exergy losses in the system are steam generator, solar field and steam turbine high pressure cylinder first stage, and the exergy loss of the three equipment in total is close to 50% of the total exergy loss. At the same time, the exergic efficiency of solar field is mainly affected by the change of direct normal irradiance. The exergic efficiency of electric heater is basically unchanged with the maximum change of 2% under different operation strategies.
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。
2023, 44(5): 1-5.
doi: 10.13832/j.jnpe.2023.05.0001
摘要：
“碳达峰、碳中和”目标的提出对我国未来能源体系发展具有深远影响。核能作为稳定的清洁能源，对于“碳达峰、碳中和”目标实现能够发挥更大作用，在发电、供热、制氢等领域均有着巨大的应用前景和需求。经过60余年发展，核能建立了完善的产业链，研发形成了“华龙一号”等具有完全自主知识产权的第三代大型商业压水堆核电技术品牌，研发了具有国际先进水平的多用途模块式小型堆“玲龙一号”，积极探索了钠冷快堆、超高温气冷堆、熔盐堆等第四代先进核能技术，持续开展聚变核能利用。同时我国核能发展也面临一些挑战，先进核能技术亟需突破。本文提出了先进核能技术的发展思路和路径，从在役核电厂智能化运行管理、三代核电批量化部署、固有安全快堆技术研发、积极研发满足高效制氢需求的超高温气冷堆、积极探索能够满足工业供热和平台供电的模块式小型堆技术、国内国际合作发展先进核能关键技术等6个方面进行了展望，为我国先进核能技术的发展给出了具体的研究目标与方向。
“碳达峰、碳中和”目标的提出对我国未来能源体系发展具有深远影响。核能作为稳定的清洁能源，对于“碳达峰、碳中和”目标实现能够发挥更大作用，在发电、供热、制氢等领域均有着巨大的应用前景和需求。经过60余年发展，核能建立了完善的产业链，研发形成了“华龙一号”等具有完全自主知识产权的第三代大型商业压水堆核电技术品牌，研发了具有国际先进水平的多用途模块式小型堆“玲龙一号”，积极探索了钠冷快堆、超高温气冷堆、熔盐堆等第四代先进核能技术，持续开展聚变核能利用。同时我国核能发展也面临一些挑战，先进核能技术亟需突破。本文提出了先进核能技术的发展思路和路径，从在役核电厂智能化运行管理、三代核电批量化部署、固有安全快堆技术研发、积极研发满足高效制氢需求的超高温气冷堆、积极探索能够满足工业供热和平台供电的模块式小型堆技术、国内国际合作发展先进核能关键技术等6个方面进行了展望，为我国先进核能技术的发展给出了具体的研究目标与方向。
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(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。
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。
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。
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。
2022, 43(1): 1-6.
doi: 10.13832/j.jnpe.2022.01.0001
Abstract:
2020, 41(4): 1-7.
Abstract:
2023, 44(1): 1-8.
doi: 10.13832/j.jnpe.2023.01.0001
Abstract:
2021, 42(4): 1-5.
doi: 10.13832/j.jnpe.2021.04.0001
Abstract:
2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
Abstract:
2022, 43(4): 1-4.
doi: 10.13832/j.jnpe.2022.04.0001
Abstract:
2019, 40(4): 1-8.
Abstract:
2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
Abstract:
2023, 44(2): 1-8.
doi: 10.13832/j.jnpe.2023.02.0001
Abstract:
2018, 39(3): 6-9.
doi: 10.13832/j.jnpe.2018.03.0006
Abstract:
