2024, 45(S2): 1-13.
doi: 10.13832/j.jnpe.2024.S2.0001
Abstract:
A digital reactor represents a comprehensive integrated platform that harnesses the power of high-precision numerical simulations, interdisciplinary data models, big data, and artificial intelligence to implement the full life-cycle management of nuclear reactors within a digital realm. To address the demands of multi-level business activities of nuclear reactors, leveraging advancements in digital technology, a digital reactor platform architecture characterized by a unified platform and two fundamental cores has been developed. Furthermore, the design of the technical architecture has been implemented through comprehensive integration and application, numerical computation cores, and big data frameworks, thereby empowering the full life-cycle management of nuclear reactors. The key technologies, problems and challenges in the application of digital reactor are analyzed and studied, and the subsequent development trend is analyzed and prospected.
A digital reactor represents a comprehensive integrated platform that harnesses the power of high-precision numerical simulations, interdisciplinary data models, big data, and artificial intelligence to implement the full life-cycle management of nuclear reactors within a digital realm. To address the demands of multi-level business activities of nuclear reactors, leveraging advancements in digital technology, a digital reactor platform architecture characterized by a unified platform and two fundamental cores has been developed. Furthermore, the design of the technical architecture has been implemented through comprehensive integration and application, numerical computation cores, and big data frameworks, thereby empowering the full life-cycle management of nuclear reactors. The key technologies, problems and challenges in the application of digital reactor are analyzed and studied, and the subsequent development trend is analyzed and prospected.
2024, 45(S2): 168-173.
doi: 10.13832/j.jnpe.2024.S2.0168
Abstract:
After the severe loss of coolant accident (LOCA), the lower head of reactor pressure vessel (RPV) will be ablated by the core melt. Therefore, it is of great significance to simulate and analyze the ablation and complex mechanical behavior of the pressure vessel wall under the condition of core melt for the design, accident prevention and mitigation of RPV. This paper firstly presents the peridynamic formulation of coupling thermo-mechanical problem with crack propagation. Then, based on the peridyanmic framework, a simple and efficient moving boundary model is proposed. The abated status of the material points is directly characterized by introducing a scalar field, which makes it unnecessary to constantly update the calculation domain in the calculation process, thus improving the calculation efficiency. Finally, the present method is used to simulate the dynamic ablation of the core melt on the pressure vessel wall and the crack propagation of the pressure vessel under the action of internal pressure. The calculation results show that there are both elastic deformation and plastic deformation in RPV under the core melt accident, and there are also complex mechanical behaviors such as damage and fracture.
After the severe loss of coolant accident (LOCA), the lower head of reactor pressure vessel (RPV) will be ablated by the core melt. Therefore, it is of great significance to simulate and analyze the ablation and complex mechanical behavior of the pressure vessel wall under the condition of core melt for the design, accident prevention and mitigation of RPV. This paper firstly presents the peridynamic formulation of coupling thermo-mechanical problem with crack propagation. Then, based on the peridyanmic framework, a simple and efficient moving boundary model is proposed. The abated status of the material points is directly characterized by introducing a scalar field, which makes it unnecessary to constantly update the calculation domain in the calculation process, thus improving the calculation efficiency. Finally, the present method is used to simulate the dynamic ablation of the core melt on the pressure vessel wall and the crack propagation of the pressure vessel under the action of internal pressure. The calculation results show that there are both elastic deformation and plastic deformation in RPV under the core melt accident, and there are also complex mechanical behaviors such as damage and fracture.
2024, 45(S2): 189-196.
doi: 10.13832/j.jnpe.2024.S2.0189
Abstract:
Following a steam generator tube rupture (SGTR) accident at a nuclear power plant, the manual adjustment control method is commonly employed, and the adjustment of the safety injection process is one of the difficulties faced by operators. For this process, the main adjustment mechanism and characteristics of the object are analyzed firstly in this paper. Combined with the experience of the operator, the automatic control structure of the safety injection control process is summarized, and the corresponding parameter setting method is designed. The intelligent elements such as adaptive PID, intelligent feedforward and reinforcement learning are integrated, so that the intelligent control strategy of the safety injection control process is innovatively proposed. In order to complete the strategy validation on the M310 full-range simulator developed by a company, this paper first develops the intelligent computing engine, then completed the control strategy configuration on the intelligent computing engine, then completes the control strategy configuration on the intelligent computing engine, and completes the real-time communication of the control signals through the MySQL. Through the simulation and test verification of different breach accidents in the simulator, the intelligent control strategies proposed in this paper are all able to realize the automatic adjustment of the injection control process, the actual cooling rate deviation is 5.89% of the cooling rate set value, the matching effect of the subcooling degree and the regulator level adjustment is good, and the performance is higher than the average of the manipulator's manual execution of the task.
Following a steam generator tube rupture (SGTR) accident at a nuclear power plant, the manual adjustment control method is commonly employed, and the adjustment of the safety injection process is one of the difficulties faced by operators. For this process, the main adjustment mechanism and characteristics of the object are analyzed firstly in this paper. Combined with the experience of the operator, the automatic control structure of the safety injection control process is summarized, and the corresponding parameter setting method is designed. The intelligent elements such as adaptive PID, intelligent feedforward and reinforcement learning are integrated, so that the intelligent control strategy of the safety injection control process is innovatively proposed. In order to complete the strategy validation on the M310 full-range simulator developed by a company, this paper first develops the intelligent computing engine, then completed the control strategy configuration on the intelligent computing engine, then completes the control strategy configuration on the intelligent computing engine, and completes the real-time communication of the control signals through the MySQL. Through the simulation and test verification of different breach accidents in the simulator, the intelligent control strategies proposed in this paper are all able to realize the automatic adjustment of the injection control process, the actual cooling rate deviation is 5.89% of the cooling rate set value, the matching effect of the subcooling degree and the regulator level adjustment is good, and the performance is higher than the average of the manipulator's manual execution of the task.
2024, 45(S2): 231-237.
doi: 10.13832/j.jnpe.2024.S2.0231
Abstract:
According to the research and development requirements of future ultra-high temperature gas cooled reactors, the design of relevant fuel elements, preparation of new coating layers, high-temperature testing and evaluation of TRISO particles before and after irradiation have been carried out. Specifically, the preparation methods of fluidized bed chemical vapor deposition for SiC coating layers, ZrC coating layers, NbC coating layers, and carbide composite coating layers are studied. And the SiC coating layer was used in the ultra-high temperature testing, including pre irradiation (up to 2500℃) and post irradiation (up to 1770℃). The research results indicate that single-phase SiC coating, ZrC coating, NbC coating, and carbide composite coating can be prepared using fluidized bed chemical vapor deposition method with liquid methyltrichlorosilane, solid ZrCl4, and NbCl5 as precursors. The gas phase carrier transportation and gas powder transportation methods are developed. The large-scale preparation of SiC coating has been successfully obtained. It was found that the SiC coating layer of TRISO particles could withstand the high temperature of 2200℃ for a short time in the high-temperature test before irradiation. At temperatures above 2100℃, partial phase transformation, grain growth, and micro decomposition could be found, but the overall coating layer remained relatively intact. The high temperature test at 1770℃ after irradiation showed that high temperatures would accelerate the diffusion of some fission elements such as Cs in the dense pyrolysis carbon layer of TRISO particles. And no damage was found in the SiC coating layer, indicating that the ability to block fission products continued to be maintained. Besides, molecular simulations were used to simulate various microstructures and mixed crystal SiC coating materials after high-temperature testing and irradiation. The above research results provide reference for the research and development and performance evaluation of fuel elements in ultra-high temperature gas-cooled reactors in China, and are of great significance to the future development of ultra-high temperature gas-cooled reactors.
According to the research and development requirements of future ultra-high temperature gas cooled reactors, the design of relevant fuel elements, preparation of new coating layers, high-temperature testing and evaluation of TRISO particles before and after irradiation have been carried out. Specifically, the preparation methods of fluidized bed chemical vapor deposition for SiC coating layers, ZrC coating layers, NbC coating layers, and carbide composite coating layers are studied. And the SiC coating layer was used in the ultra-high temperature testing, including pre irradiation (up to 2500℃) and post irradiation (up to 1770℃). The research results indicate that single-phase SiC coating, ZrC coating, NbC coating, and carbide composite coating can be prepared using fluidized bed chemical vapor deposition method with liquid methyltrichlorosilane, solid ZrCl4, and NbCl5 as precursors. The gas phase carrier transportation and gas powder transportation methods are developed. The large-scale preparation of SiC coating has been successfully obtained. It was found that the SiC coating layer of TRISO particles could withstand the high temperature of 2200℃ for a short time in the high-temperature test before irradiation. At temperatures above 2100℃, partial phase transformation, grain growth, and micro decomposition could be found, but the overall coating layer remained relatively intact. The high temperature test at 1770℃ after irradiation showed that high temperatures would accelerate the diffusion of some fission elements such as Cs in the dense pyrolysis carbon layer of TRISO particles. And no damage was found in the SiC coating layer, indicating that the ability to block fission products continued to be maintained. Besides, molecular simulations were used to simulate various microstructures and mixed crystal SiC coating materials after high-temperature testing and irradiation. The above research results provide reference for the research and development and performance evaluation of fuel elements in ultra-high temperature gas-cooled reactors in China, and are of great significance to the future development of ultra-high temperature gas-cooled reactors.
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。
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(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。
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
Abstract:
2019, 40(4): 1-8.
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2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
Abstract:
2018, 39(3): 6-9.
doi: 10.13832/j.jnpe.2018.03.0006
Abstract:
2019, 40(5): 1-5.
Abstract:
