2025, 46(S1): 1-7.
doi: 10.13832/j.jnpe.2025.S1.0001
Abstract:
This paper focuses on the irradiation test in the high-flux research reactor. Based on the Monte Carlo neutron-photon transport-burnup coupling code, the physical processes of the generation, transport, leakage, and deposition of different particles in the global core environment and the local test device environment are studied, and a high-fidelity calculation model of energy deposition that can describe the physical processes of multi loops throughout the core is established based on the repetitive geometric structure. Taking into account the dynamic changes of fuel burnup, control rod position, and their coupling relationship with the particle transport process during the operation of the high-flux research reactor, the whole reactor refueling and fuel management calculations are realized, including multi-type of fuels and burnable poisons. Validation is performed using actual critical rod position, irradiation test assembly power, and point burnup measurement values under different burnup steps. The calculation results show that the error in the core neutron effective multiplication factor (keff) is less than 1200 pcm, and the average relative errors between calculated and measured values for irradiation test assembly power and point burnup are both less than 10%, verifying the accuracy of the multi-cycle calculation method.
This paper focuses on the irradiation test in the high-flux research reactor. Based on the Monte Carlo neutron-photon transport-burnup coupling code, the physical processes of the generation, transport, leakage, and deposition of different particles in the global core environment and the local test device environment are studied, and a high-fidelity calculation model of energy deposition that can describe the physical processes of multi loops throughout the core is established based on the repetitive geometric structure. Taking into account the dynamic changes of fuel burnup, control rod position, and their coupling relationship with the particle transport process during the operation of the high-flux research reactor, the whole reactor refueling and fuel management calculations are realized, including multi-type of fuels and burnable poisons. Validation is performed using actual critical rod position, irradiation test assembly power, and point burnup measurement values under different burnup steps. The calculation results show that the error in the core neutron effective multiplication factor (keff) is less than 1200 pcm, and the average relative errors between calculated and measured values for irradiation test assembly power and point burnup are both less than 10%, verifying the accuracy of the multi-cycle calculation method.
2025, 46(S1): 82-87.
doi: 10.13832/j.jnpe.2025.S1.0082
Abstract:
With the batch construction of the independently developed third-generation pressurized water reactor nuclear power plant HPR1000 at home and abroad, China National Nuclear Corporation (CNNC) initiated the development of its follow-up reactor (AHPR1000) in 2019 to further enhance the safety, economy, advancement, operational reliability, environmental friendliness, and intelligence level of HPR1000. The design innovation of safety systems and facilities is a core part of this development. The safety systems and facilities of the AHPR1000 reactor are mainly used to ensure reactor safety and control/contain radioactive material releases under accident conditions. This paper proposes a design concept combining "passive and active" systems for the safety system configuration of AHPR1000. Starting from the top-level safety philosophy and design principles, it focuses on introducing the safety functions, safety configuration, accident response strategies, and experimental validation.
With the batch construction of the independently developed third-generation pressurized water reactor nuclear power plant HPR1000 at home and abroad, China National Nuclear Corporation (CNNC) initiated the development of its follow-up reactor (AHPR1000) in 2019 to further enhance the safety, economy, advancement, operational reliability, environmental friendliness, and intelligence level of HPR1000. The design innovation of safety systems and facilities is a core part of this development. The safety systems and facilities of the AHPR1000 reactor are mainly used to ensure reactor safety and control/contain radioactive material releases under accident conditions. This paper proposes a design concept combining "passive and active" systems for the safety system configuration of AHPR1000. Starting from the top-level safety philosophy and design principles, it focuses on introducing the safety functions, safety configuration, accident response strategies, and experimental validation.
2025, 46(S1): 166-180.
doi: 10.13832/j.jnpe.2025.S1.0166
Abstract:
The discrete ordinates (SN) method is one of the primary methods for calculating fast neutron flux in reactor pressure vessels (RPVs). The three-dimensional neutron flux synthesis method based on two-dimensional plus one-dimensional SN calculations (referred to as the synthesis method) has higher computational efficiency compared with the direct three-dimensional discrete ordinates method (referred to as the three-dimensional calculation method). However, the approximations in the source and geometry processing can affect the accuracy of the synthesis method. In order to analyze the impact of source and geometric approximations on the synthesis method, a benchmark model suitable for the synthesis method was established. The relative error of fast neutron flux calculated by the synthesis method and the three-dimensional calculation of the benchmark model is used as a reference to analyze the influence of source and geometric factors on the synthesis method. In the analysis of source effects, non-uniform radial, axial, and azimuthal power distributions are introduced into the benchmark model to examine the changes in the relative error of fast neutron flux between synthetic calculations and three-dimensional calculations. In the analysis of geometric effects, the core structure of the benchmark model is altered to a square core and a stepped core, respectively, to sequentially analyze the changes in the relative error of fast neutron flux between synthetic and three-dimensional calculations. The results indicate that the maximum relative errors caused by the radial and axial power distributions are both within 1.5%, while the azimuthal power distribution leads to a relative error of 3.5% at the reactor cavity. The square and stepped core structures result in relative errors of 20% and 22%, respectively, in the cavity. In the computations for the typical pressurized water reactor HBR-2, the relative error between the synthesis method and the three-dimensional fast neutron flux calculations at the cavity was 11.65%. These findings demonstrate that the accuracy of the synthesis method for the reactor cavity region still requires improvement.
The discrete ordinates (SN) method is one of the primary methods for calculating fast neutron flux in reactor pressure vessels (RPVs). The three-dimensional neutron flux synthesis method based on two-dimensional plus one-dimensional SN calculations (referred to as the synthesis method) has higher computational efficiency compared with the direct three-dimensional discrete ordinates method (referred to as the three-dimensional calculation method). However, the approximations in the source and geometry processing can affect the accuracy of the synthesis method. In order to analyze the impact of source and geometric approximations on the synthesis method, a benchmark model suitable for the synthesis method was established. The relative error of fast neutron flux calculated by the synthesis method and the three-dimensional calculation of the benchmark model is used as a reference to analyze the influence of source and geometric factors on the synthesis method. In the analysis of source effects, non-uniform radial, axial, and azimuthal power distributions are introduced into the benchmark model to examine the changes in the relative error of fast neutron flux between synthetic calculations and three-dimensional calculations. In the analysis of geometric effects, the core structure of the benchmark model is altered to a square core and a stepped core, respectively, to sequentially analyze the changes in the relative error of fast neutron flux between synthetic and three-dimensional calculations. The results indicate that the maximum relative errors caused by the radial and axial power distributions are both within 1.5%, while the azimuthal power distribution leads to a relative error of 3.5% at the reactor cavity. The square and stepped core structures result in relative errors of 20% and 22%, respectively, in the cavity. In the computations for the typical pressurized water reactor HBR-2, the relative error between the synthesis method and the three-dimensional fast neutron flux calculations at the cavity was 11.65%. These findings demonstrate that the accuracy of the synthesis method for the reactor cavity region still requires improvement.
2025, 46(S1): 207-212.
doi: 10.13832/j.jnpe.2025.S1.0207
Abstract:
The heat source (power supply) of nuclear fission reactor is expected to be applied to deep space exploration, catalog scientific research station, interstellar navigation and other fields in the future due to its many advantages, such as no demand for solar orientation, small impact from space environment, large power range and large regulation range. Based on the fourth generation nuclear reactor technology—molten salt reactor, this paper proposes a conceptual design for a 40 kW space nuclear reactor utilizing liquid molten salt as nuclear fuel coupled with heat pipe cooling. A new reactivity control strategy combining control drum (for power regulation) and safety drum (for emergency shutdown) is proposed. The physical model of liquid molten salt space nuclear reactor is established, and the key physical characteristics of the core, such as neutron energy spectrum, neutron flux distribution, temperature effect and burnup depth, are obtained based on the Monte Carlo code MCNP and RMC analysis. The influence of control drum angle on reactivity and the reactivity control and core safety under partial failure of dual drums are further analyzed. The results show that the liquid molten salt space nuclear reactor designed in this paper can operate at full power for 10 years. The control drum arrangement can meet the core safety requirements under the failure of some control drums. This research can provide design reference for the control strategy of space liquid molten salt reactor and space reactor.
The heat source (power supply) of nuclear fission reactor is expected to be applied to deep space exploration, catalog scientific research station, interstellar navigation and other fields in the future due to its many advantages, such as no demand for solar orientation, small impact from space environment, large power range and large regulation range. Based on the fourth generation nuclear reactor technology—molten salt reactor, this paper proposes a conceptual design for a 40 kW space nuclear reactor utilizing liquid molten salt as nuclear fuel coupled with heat pipe cooling. A new reactivity control strategy combining control drum (for power regulation) and safety drum (for emergency shutdown) is proposed. The physical model of liquid molten salt space nuclear reactor is established, and the key physical characteristics of the core, such as neutron energy spectrum, neutron flux distribution, temperature effect and burnup depth, are obtained based on the Monte Carlo code MCNP and RMC analysis. The influence of control drum angle on reactivity and the reactivity control and core safety under partial failure of dual drums are further analyzed. The results show that the liquid molten salt space nuclear reactor designed in this paper can operate at full power for 10 years. The control drum arrangement can meet the core safety requirements under the failure of some control drums. This research can provide design reference for the control strategy of space liquid molten salt reactor and space reactor.
2025, 46(S1): 242-249.
doi: 10.13832/j.jnpe.2025.S1.0242
Abstract:
In order to meet the numerical simulation of various operating conditions in nuclear reactors, it is necessary to provide cross-section data at different temperature. However, the cross sections provided in the evaluated nuclear data library are the data at 0 K. Therefore, in order to meet the needs of core physics numerical simulation calculations, it is necessary to perform Doppler broadening processing based on the evaluated nuclear data in the evaluated nuclear data library to obtain continuous energy pointwise cross sections at different temperatures. Through the Kernel Broadening accurate Doppler broadening calculation method, the theoretical derivation and code development of the Doppler broadening calculation method have been completed. Based on the CENDL-3.2 evaluated nuclear data library, the broadening cross sections at different temperatures are calculated based on the doppler_broad module developed in this paper and the BROADR module in the NJOY2016, respectively, and a computational analysis is performed to verify the rationality of some convergence parameters in the Doppler broadening calculation process. The numerical results indicate that for the four temperature points of 293.6 K, 600 K, 900 K and 108 K, the calculation results in this paper are in good agreement with the NJOY2016 calculation results. The value of the NMAX parameter, which determines the limit of adjacent broadening energy points, has a significant impact on the broadening cross sections. The maximum relative deviation of the broadening cross sections for different values is 1.091%.
In order to meet the numerical simulation of various operating conditions in nuclear reactors, it is necessary to provide cross-section data at different temperature. However, the cross sections provided in the evaluated nuclear data library are the data at 0 K. Therefore, in order to meet the needs of core physics numerical simulation calculations, it is necessary to perform Doppler broadening processing based on the evaluated nuclear data in the evaluated nuclear data library to obtain continuous energy pointwise cross sections at different temperatures. Through the Kernel Broadening accurate Doppler broadening calculation method, the theoretical derivation and code development of the Doppler broadening calculation method have been completed. Based on the CENDL-3.2 evaluated nuclear data library, the broadening cross sections at different temperatures are calculated based on the doppler_broad module developed in this paper and the BROADR module in the NJOY2016, respectively, and a computational analysis is performed to verify the rationality of some convergence parameters in the Doppler broadening calculation process. The numerical results indicate that for the four temperature points of 293.6 K, 600 K, 900 K and 108 K, the calculation results in this paper are in good agreement with the NJOY2016 calculation results. The value of the NMAX parameter, which determines the limit of adjacent broadening energy points, has a significant impact on the broadening cross sections. The maximum relative deviation of the broadening cross sections for different values is 1.091%.
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:
