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doi: 10.13832/j.jnpe.2025.05.0205
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doi: 10.13832/j.jnpe.2025.06.0274
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doi: 10.13832/j.jnpe.2025.07.0345
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doi: 10.13832/j.jnpe.2025.05.0230
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doi: 10.13832/j.jnpe.2025.03.0096
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2025, 46(5): 1-11.
doi: 10.13832/j.jnpe.2025.06.0260
Abstract:
The construction of a solution-type medical isotope test reactor for the production of isotopes such as 99Mo and 131I is one of the important initiatives to achieve self-sufficient and controllable supply in China's medical isotope market. This article briefly introduces the application status, production principles, and production methods of medical isotopes, as well as the development overview of homogeneous solution-type reactors both domestically and internationally. It systematically elaborates on the system composition and design of the medical isotope test reactor, including the reactor and its major systems, isotope extraction process systems, and supporting systems. Furthermore, it provides a detailed explanation of the main technical issues, such as reactivity stability, radiation protection design, prevention of fuel solution precipitation, corrosion resistance of structural materials, critical safety of fuel solutions, isotope extraction processes, uranium recovery technology, fuel purification technology, and radioactive waste gas treatment technology.
The construction of a solution-type medical isotope test reactor for the production of isotopes such as 99Mo and 131I is one of the important initiatives to achieve self-sufficient and controllable supply in China's medical isotope market. This article briefly introduces the application status, production principles, and production methods of medical isotopes, as well as the development overview of homogeneous solution-type reactors both domestically and internationally. It systematically elaborates on the system composition and design of the medical isotope test reactor, including the reactor and its major systems, isotope extraction process systems, and supporting systems. Furthermore, it provides a detailed explanation of the main technical issues, such as reactivity stability, radiation protection design, prevention of fuel solution precipitation, corrosion resistance of structural materials, critical safety of fuel solutions, isotope extraction processes, uranium recovery technology, fuel purification technology, and radioactive waste gas treatment technology.
2025, 46(5): 12-21.
doi: 10.13832/j.jnpe.2024.090020
Abstract:
Liquid lithium is widely utilized as a metal coolant in space fast reactors due to its smaller neutron absorption cross-section, lower density, and superior thermophysical properties, resulting in reduced core mass and enhanced heat transfer efficiency. However, under neutron irradiation, lithium reacts with neutrons to produce helium gas. The accumulation of helium increases thermal resistance in affected regions and reduces the heat transfer efficiency of system equipment. In this study, the Monte Carlo code OpenMC is employed to conduct helium production calculations for three lithium-cooled space fast reactor design models featuring different fuel space separations. The aim is to analyze the impact of these separations on the helium production behavior within the reactor core. The paper also calculates the total helium yield of the core, analyzes the helium production capacity of various nuclear reactions between liquid lithium and neutrons, and examines helium production under different 7Li enrichments, and the accurate calculation of helium production in the core is realized. Focusing on changes in helium production as a function of core burnup, the results indicate that larger fuel space separations result in higher helium production. Additionally, increasing 7Li enrichment significantly reduces helium production, with a maximum 68.76% reduction observed in helium yield when enrichment is elevated from 95% to 99%. The paper provides valuable insights for optimizing lithium-cooled space fast reactors.
Liquid lithium is widely utilized as a metal coolant in space fast reactors due to its smaller neutron absorption cross-section, lower density, and superior thermophysical properties, resulting in reduced core mass and enhanced heat transfer efficiency. However, under neutron irradiation, lithium reacts with neutrons to produce helium gas. The accumulation of helium increases thermal resistance in affected regions and reduces the heat transfer efficiency of system equipment. In this study, the Monte Carlo code OpenMC is employed to conduct helium production calculations for three lithium-cooled space fast reactor design models featuring different fuel space separations. The aim is to analyze the impact of these separations on the helium production behavior within the reactor core. The paper also calculates the total helium yield of the core, analyzes the helium production capacity of various nuclear reactions between liquid lithium and neutrons, and examines helium production under different 7Li enrichments, and the accurate calculation of helium production in the core is realized. Focusing on changes in helium production as a function of core burnup, the results indicate that larger fuel space separations result in higher helium production. Additionally, increasing 7Li enrichment significantly reduces helium production, with a maximum 68.76% reduction observed in helium yield when enrichment is elevated from 95% to 99%. The paper provides valuable insights for optimizing lithium-cooled space fast reactors.
2025, 46(5): 132-138.
doi: 10.13832/j.jnpe.2024.090032
Abstract:
During the in-service inspection of the bimetallic weld in the low-pressure safety injection system pipeline of a nuclear power unit, ductility dip cracking (DDC) was identified on the overlay weld surface of ЭА-395/9 welding material. Through research on the hot cracking sensitivity of the ЭА-395/9 welding material, analysis of welding stress, and experimental study of the welding process, it was determined that when the overlay welding heat input is ≥ 16.3 kJ/cm, only 5% strain is sufficient to initiate cracking and form DDC. When the welding heat input of ЭА-395/9 welding increases, the columnar crystals become coarse and chain like precipitates form at the grain boundaries, resulting in high welding stress and providing metallurgical and mechanical conditions for the occurrence of DDC. By using a welding current of 120~130A, reducing the arc lateral swing amplitude, and controlling the interlayer temperature below 100℃, it is possible to effectively prevent the formation of DDC in the overlayweld of ЭА-395/9 welding material.
During the in-service inspection of the bimetallic weld in the low-pressure safety injection system pipeline of a nuclear power unit, ductility dip cracking (DDC) was identified on the overlay weld surface of ЭА-395/9 welding material. Through research on the hot cracking sensitivity of the ЭА-395/9 welding material, analysis of welding stress, and experimental study of the welding process, it was determined that when the overlay welding heat input is ≥ 16.3 kJ/cm, only 5% strain is sufficient to initiate cracking and form DDC. When the welding heat input of ЭА-395/9 welding increases, the columnar crystals become coarse and chain like precipitates form at the grain boundaries, resulting in high welding stress and providing metallurgical and mechanical conditions for the occurrence of DDC. By using a welding current of 120~130A, reducing the arc lateral swing amplitude, and controlling the interlayer temperature below 100℃, it is possible to effectively prevent the formation of DDC in the overlayweld of ЭА-395/9 welding material.
2025, 46(5): 171-179.
doi: 10.13832/j.jnpe.2024.090030
Abstract:
To ensure the operation safety of low-temperature heating reactors, it is crucial to study reactor power control strategies that meet their operational requirements. To this end, this study takes the low-temperature heating reactor coolant system as the research object and proposes two control strategies: a dual-feedback control strategy and a cascade control strategy. Their respective effects on the coupled control of reactor power and core coolant outlet temperature are investigated. The simulation results show that both control strategies can effectively control the low-temperature heating reactor coolant system under step and linear load change conditions. Under the cascade control strategy, the coolant outlet temperature is the main controlled variable, whose change magnitude is small in the load change conditions, but the settling time of reactor power is prolonged. Therefore, the cascade control strategy is more suitable for the linear load change conditions. Under the double-feedback control strategy, the reactor power control channel and the temperature control channel are in a parallel configuration, allowing both control performances to be taken into account. Therefore, the double-feedback control strategy is more suitable for the step load change conditions. This study can provide a reference for the development and optimization of low-temperature heating reactor coolant system control strategy.
To ensure the operation safety of low-temperature heating reactors, it is crucial to study reactor power control strategies that meet their operational requirements. To this end, this study takes the low-temperature heating reactor coolant system as the research object and proposes two control strategies: a dual-feedback control strategy and a cascade control strategy. Their respective effects on the coupled control of reactor power and core coolant outlet temperature are investigated. The simulation results show that both control strategies can effectively control the low-temperature heating reactor coolant system under step and linear load change conditions. Under the cascade control strategy, the coolant outlet temperature is the main controlled variable, whose change magnitude is small in the load change conditions, but the settling time of reactor power is prolonged. Therefore, the cascade control strategy is more suitable for the linear load change conditions. Under the double-feedback control strategy, the reactor power control channel and the temperature control channel are in a parallel configuration, allowing both control performances to be taken into account. Therefore, the double-feedback control strategy is more suitable for the step load change conditions. This study can provide a reference for the development and optimization of low-temperature heating reactor coolant system control strategy.
2025, 46(5): 205-210.
doi: 10.13832/j.jnpe.2025.04.0184
Abstract:
Nuclear power is a safe, economical, and efficient clean energy source that serves as a strong support for China's national strategy of "Carbon Peaking and Carbon Neutrality". Nuclear power equipment is the core enabler of nuclear power plants' critical functions. This paper analyzes the development status of nuclear power in China and elaborates on the current state of nuclear power equipment. It examines development opportunities in two major directions: power generation and multi-purpose applications. The challenges faced by nuclear power equipment are identified, including performance improvements of existing equipment, R&D of equipment for advanced reactor types, and the establishment and transformation of the industrial chain. Development suggestions are provided, which may serve as a reference for the high-quality development of China's nuclear power equipment.
Nuclear power is a safe, economical, and efficient clean energy source that serves as a strong support for China's national strategy of "Carbon Peaking and Carbon Neutrality". Nuclear power equipment is the core enabler of nuclear power plants' critical functions. This paper analyzes the development status of nuclear power in China and elaborates on the current state of nuclear power equipment. It examines development opportunities in two major directions: power generation and multi-purpose applications. The challenges faced by nuclear power equipment are identified, including performance improvements of existing equipment, R&D of equipment for advanced reactor types, and the establishment and transformation of the industrial chain. Development suggestions are provided, which may serve as a reference for the high-quality development of China's nuclear power equipment.
2025, 46(5): 217-223.
doi: 10.13832/j.jnpe.2024.10.0049
Abstract:
The loose parts carried by the coolant in the reactor primary circuit can move to the lower grating plate, causing collisions with the lower grating plate and further blocking the diversion hole. After being transmitted to the top cover of the reactor pressure vessel through internal structures, the collision mechanical noise of the lower grating plate experiences signal attenuation and is masked by the reactor background noise, making direct identification impossible. Therefore, this study first conducted simulation tests to obtain background noise data and weak collision mechanical noise data of the lower grating plate. Then, wavelet packet decomposition (WPD) combined with a kurtosis threshold method was used to denoise the weak collision mechanical noise submerged in background noise. Finally, based on the denoised collision signal, loose parts collision identification was performed, and a code for identifying collision events of the lower grating plate in the reactor was developed. Test results show that the proposed denoising method is effective, and the developed code can effectively identify the signals of lower grating plate collision events submerged in background noise.
The loose parts carried by the coolant in the reactor primary circuit can move to the lower grating plate, causing collisions with the lower grating plate and further blocking the diversion hole. After being transmitted to the top cover of the reactor pressure vessel through internal structures, the collision mechanical noise of the lower grating plate experiences signal attenuation and is masked by the reactor background noise, making direct identification impossible. Therefore, this study first conducted simulation tests to obtain background noise data and weak collision mechanical noise data of the lower grating plate. Then, wavelet packet decomposition (WPD) combined with a kurtosis threshold method was used to denoise the weak collision mechanical noise submerged in background noise. Finally, based on the denoised collision signal, loose parts collision identification was performed, and a code for identifying collision events of the lower grating plate in the reactor was developed. Test results show that the proposed denoising method is effective, and the developed code can effectively identify the signals of lower grating plate collision events submerged in background noise.
2025, 46(5): 234-242.
doi: 10.13832/j.jnpe.2024.090019
Abstract:
A novel training dataset processing method with high robustness was proposed to address the poor robustness in data-driven analytic redundant sensor models, and an sensor online monitoring and self-correction method was developed based on a auto-associative multivariate Long Short-Term Memory (LSTM) artificial neural network model. The method was validated using actual sensor measurement data retrieved from a pressurized water reactor engineering test facility. The results indicate that this research method can achieve high-precision and robust reconstruction of sensor signals, hence meets the requirements of online monitoring and robust self-correction of nuclear reactor sensors.
A novel training dataset processing method with high robustness was proposed to address the poor robustness in data-driven analytic redundant sensor models, and an sensor online monitoring and self-correction method was developed based on a auto-associative multivariate Long Short-Term Memory (LSTM) artificial neural network model. The method was validated using actual sensor measurement data retrieved from a pressurized water reactor engineering test facility. The results indicate that this research method can achieve high-precision and robust reconstruction of sensor signals, hence meets the requirements of online monitoring and robust self-correction of nuclear reactor sensors.
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。
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。
2016, 37(3): 34-38.
doi: 10.13832/j.jnpe.2016.03.0034
摘要：
基于热力学第一定律,开展超临界二氧化碳（S-CO2）布雷顿循环的热力学特性研究。在设备模型构建和初始条件假设的基础上,研究系统采用再压缩循环的热力学特性和参数限制条件。针对进出口温差较大的热源系统,提出了复叠式分流循环方案,开展热力学特性分析和评价,并与再压缩循环进行定量比较,得出各自的适用对象。
基于热力学第一定律,开展超临界二氧化碳（S-CO2）布雷顿循环的热力学特性研究。在设备模型构建和初始条件假设的基础上,研究系统采用再压缩循环的热力学特性和参数限制条件。针对进出口温差较大的热源系统,提出了复叠式分流循环方案,开展热力学特性分析和评价,并与再压缩循环进行定量比较,得出各自的适用对象。
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。
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。
2023, 44(S2): 33-38.
doi: 10.13832/j.jnpe.2023.S2.0033
摘要：
针对传统深度学习方法对滚动轴承剩余使用寿命（RUL）预测准确性不高的问题，提出一种基于注意力机制的卷积神经网络和双向长短期记忆网络的混合RUL模型（CNN-BiLSTM-AM），并运用该混合模型对滚动轴承的RUL进行预测。首先通过轴承原始振动信号的峭度特征确定轴承首次预测时间（FPT）；其次对FPT后的原始振动信号进行降噪、归一化处理并通过CNN-BiLSTM-AM混合模型对2种不同工况下的滚动轴承的RUL进行预测；最后，将CNN-BiLSTM-AM混合模型与几种传统模型进行对比。结果表明，CNN-BiLSTM-AM混合模型对滚动轴承的RUL更为有效，并具有泛化性能。
针对传统深度学习方法对滚动轴承剩余使用寿命（RUL）预测准确性不高的问题，提出一种基于注意力机制的卷积神经网络和双向长短期记忆网络的混合RUL模型（CNN-BiLSTM-AM），并运用该混合模型对滚动轴承的RUL进行预测。首先通过轴承原始振动信号的峭度特征确定轴承首次预测时间（FPT）；其次对FPT后的原始振动信号进行降噪、归一化处理并通过CNN-BiLSTM-AM混合模型对2种不同工况下的滚动轴承的RUL进行预测；最后，将CNN-BiLSTM-AM混合模型与几种传统模型进行对比。结果表明，CNN-BiLSTM-AM混合模型对滚动轴承的RUL更为有效，并具有泛化性能。
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(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:
2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
Abstract:
2019, 40(4): 1-8.
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:
