2025, 46(4): 1-9.
doi: 10.13832/j.jnpe.2025.05.0206
Abstract:
The reactor core is a critical component of nuclear power systems with a complex geometric structure, and it experiences strong coupling effects between various physical fields. High-precision thermal-hydraulic and multi-physics coupling analysis of the core is essential for ensuring the design and safety analysis of advanced nuclear power systems. The Nuclear Reactor Thermal-Hydraulic Laboratory (NuTHeL) at Xi'an Jiaotong University has built a core-heat-flow-deposition multi-physics coupling analysis model, and has independently developed the CorTAF Three-Dimensional Cross-Scale Multi-Physics Coupling Analysis Code for Nuclear Reactor Core, which allows CFD-based multi-physics calculations and predictions for the entire pressure vessel. Validation and verification work has also been conducted based on international benchmark problems. In recent years, the research team has continually developed and refined the mathematical and physical models of the code. Currently, the CorTAF code supports cross-scale coupling calculations for multiple reactor types (PWR, LFR, SFR), physical fields (Neutronics, Thermal hydraulic, Deposition), and system structures (Core, Lower plenum, Upper plenum). This paper reviews the development process of the CorTAF series codes, presents their main functions and applications in PWR calculation, summarizes the current computational results, and discusses the future direction of the program's development.
The reactor core is a critical component of nuclear power systems with a complex geometric structure, and it experiences strong coupling effects between various physical fields. High-precision thermal-hydraulic and multi-physics coupling analysis of the core is essential for ensuring the design and safety analysis of advanced nuclear power systems. The Nuclear Reactor Thermal-Hydraulic Laboratory (NuTHeL) at Xi'an Jiaotong University has built a core-heat-flow-deposition multi-physics coupling analysis model, and has independently developed the CorTAF Three-Dimensional Cross-Scale Multi-Physics Coupling Analysis Code for Nuclear Reactor Core, which allows CFD-based multi-physics calculations and predictions for the entire pressure vessel. Validation and verification work has also been conducted based on international benchmark problems. In recent years, the research team has continually developed and refined the mathematical and physical models of the code. Currently, the CorTAF code supports cross-scale coupling calculations for multiple reactor types (PWR, LFR, SFR), physical fields (Neutronics, Thermal hydraulic, Deposition), and system structures (Core, Lower plenum, Upper plenum). This paper reviews the development process of the CorTAF series codes, presents their main functions and applications in PWR calculation, summarizes the current computational results, and discusses the future direction of the program's development.
2025, 46(4): 10-17.
doi: 10.13832/j.jnpe.2024.070068
Abstract:
To address the deficiencies in neutronics calculations of the Four Petal-shaped Helical Fuel Rod (FPHF) and further determine the influence of the geometric characteristics of the FPHF on its neutron behavior, this paper uses DAG-OpenMC to construct an accurate neutronics calculation model of the FPHF. The study examines the impact of the FPHF's geometric characteristics on neutron behavior from three aspects: fuel rod diameter, cross-sectional shape, and helix angle. The fuel rod diameters are set to 3.5 mm, 6.3 mm, and 9.5 mm; the ratio of the concave arc to the convex arc ranges from 0.1 to 3.0; and the helix angle ranges from 360° to 1080°. The results show that when the fuel rod diameter increases from 3.5 mm to 9.5 mm, the radial power peak factor of the FPHF increases by 5.15%, and the non-uniformity of the neutron flux distribution rises. When the ratio of the concave arc to the convex arc increases from 0.1 to 3.0, the fission reaction rate Rf decreases by 0.19%, and the effective multiplication factor drops by 441.5 pcm (1pcm=10–5). The influence of the helix angle on the moderation effect of the fuel rod and the radial flux distribution is negligible. Therefore, except for the helix angle of the fuel rod, both the fuel rod diameter and cross-sectional shape studied in this paper have significant effects on the neutronics characteristics of the FPHF.
To address the deficiencies in neutronics calculations of the Four Petal-shaped Helical Fuel Rod (FPHF) and further determine the influence of the geometric characteristics of the FPHF on its neutron behavior, this paper uses DAG-OpenMC to construct an accurate neutronics calculation model of the FPHF. The study examines the impact of the FPHF's geometric characteristics on neutron behavior from three aspects: fuel rod diameter, cross-sectional shape, and helix angle. The fuel rod diameters are set to 3.5 mm, 6.3 mm, and 9.5 mm; the ratio of the concave arc to the convex arc ranges from 0.1 to 3.0; and the helix angle ranges from 360° to 1080°. The results show that when the fuel rod diameter increases from 3.5 mm to 9.5 mm, the radial power peak factor of the FPHF increases by 5.15%, and the non-uniformity of the neutron flux distribution rises. When the ratio of the concave arc to the convex arc increases from 0.1 to 3.0, the fission reaction rate Rf decreases by 0.19%, and the effective multiplication factor drops by 441.5 pcm (1pcm=10–5). The influence of the helix angle on the moderation effect of the fuel rod and the radial flux distribution is negligible. Therefore, except for the helix angle of the fuel rod, both the fuel rod diameter and cross-sectional shape studied in this paper have significant effects on the neutronics characteristics of the FPHF.
2025, 46(4): 49-59.
doi: 10.13832/j.jnpe.2024.070037
Abstract:
Due to the lack of reliable empirical correlations and related research results for predicting the flow and heat transfer of superheated steam in rod bundle channels at low Reynolds numbers, numerical simulation method is used to study superheated steam in rod bundle channels at low Reynolds numbers (Rein=1937.90~9471.24). This method is based on the Large Eddy Simulation (LES) turbulence model to investigate the effects of inlet steam velocity, degree of superheat, initial wall temperature, outlet steam pressure, and grid diameter ratio on the convective heat transfer characteristics of superheated steam in rod bundle channels at low Reynolds numbers, and to further modify the current empirical correlations. The numerical simulation results showed that the increase in inlet steam velocity, degree of superheat, initial wall temperature, outlet steam pressure, and grid diameter ratio all increased the convective heat transfer coefficient. As the inlet steam velocity, outlet steam pressure, and grid diameter ratio increased, the Nusselt number increased. As the inlet steam degree of superheat and initial wall temperature increased, the Nusselt number decreased. The error of the modified Dittus-Boelter empirical correlation was within 10%, providing a basis for guiding practical engineering applications and ensuring the safety of pressurized water reactor cores.
Due to the lack of reliable empirical correlations and related research results for predicting the flow and heat transfer of superheated steam in rod bundle channels at low Reynolds numbers, numerical simulation method is used to study superheated steam in rod bundle channels at low Reynolds numbers (Rein=1937.90~9471.24). This method is based on the Large Eddy Simulation (LES) turbulence model to investigate the effects of inlet steam velocity, degree of superheat, initial wall temperature, outlet steam pressure, and grid diameter ratio on the convective heat transfer characteristics of superheated steam in rod bundle channels at low Reynolds numbers, and to further modify the current empirical correlations. The numerical simulation results showed that the increase in inlet steam velocity, degree of superheat, initial wall temperature, outlet steam pressure, and grid diameter ratio all increased the convective heat transfer coefficient. As the inlet steam velocity, outlet steam pressure, and grid diameter ratio increased, the Nusselt number increased. As the inlet steam degree of superheat and initial wall temperature increased, the Nusselt number decreased. The error of the modified Dittus-Boelter empirical correlation was within 10%, providing a basis for guiding practical engineering applications and ensuring the safety of pressurized water reactor cores.
2025, 46(4): 117-124.
doi: 10.13832/j.jnpe.2024.070038
Abstract:
Helical cruciform fuel (HCF) has a complex geometry, which poses a higher challenge to the study of burnup characteristics. The traditional concentric circle burnup region division method could not accurately simulate the different burnup at different positions caused by the complex geometry of HCF in the burnup, and lack corresponding three-dimensional refined numerical analysis method to predict the burnup characteristics. In this paper, a hexahedral burnup region division and Computer-Aided Design (CAD) geometric modeling method are proposed for HCF. By taking the slice, minimum twist unit and single fuel of HCF as the research object, three-dimensional burnup calculation are realized, and variable distribution under different burnups, nuclear density and nuclear reaction rate of typical nuclides 235U, 238U and 239Pu at concave and convex position are obtained. The results show that the fast neutron flux, thermal neutron flux and power density distribution of radial circumferential HCF show great non-uniformity. The circumferential non-uniformity increases with the depletion of fuel, the burnup in the convex position is 15.92 MW·d/kg deeper than that in the concave. The influence of axial twist on the physical variables of the convex position of fuel is greater than that of the concave position. Three-dimensional refined analysis of burnup characteristics provides a basis for high-fidelity coupling calculation of neutronic physics, thermal-hydraulics and mechanics of HCF.
Helical cruciform fuel (HCF) has a complex geometry, which poses a higher challenge to the study of burnup characteristics. The traditional concentric circle burnup region division method could not accurately simulate the different burnup at different positions caused by the complex geometry of HCF in the burnup, and lack corresponding three-dimensional refined numerical analysis method to predict the burnup characteristics. In this paper, a hexahedral burnup region division and Computer-Aided Design (CAD) geometric modeling method are proposed for HCF. By taking the slice, minimum twist unit and single fuel of HCF as the research object, three-dimensional burnup calculation are realized, and variable distribution under different burnups, nuclear density and nuclear reaction rate of typical nuclides 235U, 238U and 239Pu at concave and convex position are obtained. The results show that the fast neutron flux, thermal neutron flux and power density distribution of radial circumferential HCF show great non-uniformity. The circumferential non-uniformity increases with the depletion of fuel, the burnup in the convex position is 15.92 MW·d/kg deeper than that in the concave. The influence of axial twist on the physical variables of the convex position of fuel is greater than that of the concave position. Three-dimensional refined analysis of burnup characteristics provides a basis for high-fidelity coupling calculation of neutronic physics, thermal-hydraulics and mechanics of HCF.
2025, 46(4): 159-167.
doi: 10.13832/j.jnpe.2024.080051
Abstract:
At present, there are two seismic categories for nuclear power plant equipment, the first (high) is called seismic category I, and the second is called non-seismic category I. The seismic design and seismic test of seismic category I are quite mature, and there are relevant national standards, while there are no relevant standards to follow for non-seismic category I equipment. In order to put forward more reasonable seismic requirements for some non-seismic category I but functionally important equipment that does not require safety shutdown earthquake resistance, this study proposes to classify the non-seismic category I equipment into the conventional important seismic category and the conventional general seismic category. In this study, the corresponding peak ground horizontal acceleration and required ground horizontal response spectra are given according to the national standard spectrum for the conventional important seismic category and the conventional general seismic category. The earthquake fortification for conventional important seismic category is increased from 50-year exceedance probability of 10% to 2%. The results show that the seismic fortifications standards for conventional important seismic category are to be in line with the international advanced standards and meet the new requirements of the National Nuclear Safety Administration for the emergency center of nuclear power plants after the Fukushima accident and the national standard GB 50260-2013. Therefore, the seismic categories suggested in this study are applicable to the seismic design of nuclear power plants in China.
At present, there are two seismic categories for nuclear power plant equipment, the first (high) is called seismic category I, and the second is called non-seismic category I. The seismic design and seismic test of seismic category I are quite mature, and there are relevant national standards, while there are no relevant standards to follow for non-seismic category I equipment. In order to put forward more reasonable seismic requirements for some non-seismic category I but functionally important equipment that does not require safety shutdown earthquake resistance, this study proposes to classify the non-seismic category I equipment into the conventional important seismic category and the conventional general seismic category. In this study, the corresponding peak ground horizontal acceleration and required ground horizontal response spectra are given according to the national standard spectrum for the conventional important seismic category and the conventional general seismic category. The earthquake fortification for conventional important seismic category is increased from 50-year exceedance probability of 10% to 2%. The results show that the seismic fortifications standards for conventional important seismic category are to be in line with the international advanced standards and meet the new requirements of the National Nuclear Safety Administration for the emergency center of nuclear power plants after the Fukushima accident and the national standard GB 50260-2013. Therefore, the seismic categories suggested in this study are applicable to the seismic design of nuclear power plants in China.
2025, 46(4): 192-198.
doi: 10.13832/j.jnpe.2024.080017
Abstract:
Accurate prediction is fundamental to the condition monitoring and operational maintenance of nuclear power plants (NPPs). To improve the dynamic trend prediction of systems and components, this paper proposes a time series prediction method based on signal decomposition strategy. Firstly, the original time series signal is decomposed into two subsequences containing high-frequency noise and low-frequency trend respectively, utilizing variational mode decomposition (VMD). Then, the gated recurrent unit optimized by the Bayesian optimization algorithm (BOA-GRU) and autoregressive integrated moving average (ARIMA) are employed to forecast these high-frequency and low-frequency subsequences separately. Finally, the predicted values of both subsequences are recombined to derive the forecast of the original signal. The proposed hybrid model is applied to perform single-step and multi-step predictions on the time series signals from the reactor coolant pump of a specific NPP, and the prediction accuracy is evaluated using metrics such as root mean square error (RMSE), mean absolute percentage error (MAPE), and mean absolute error (MAE). The results demonstrate that the proposed hybrid model can accurately predict and track the operational status of main coolant pump, and comparisons with baseline models highlight the advantages of the proposed hybrid model in complex signal prediction.
Accurate prediction is fundamental to the condition monitoring and operational maintenance of nuclear power plants (NPPs). To improve the dynamic trend prediction of systems and components, this paper proposes a time series prediction method based on signal decomposition strategy. Firstly, the original time series signal is decomposed into two subsequences containing high-frequency noise and low-frequency trend respectively, utilizing variational mode decomposition (VMD). Then, the gated recurrent unit optimized by the Bayesian optimization algorithm (BOA-GRU) and autoregressive integrated moving average (ARIMA) are employed to forecast these high-frequency and low-frequency subsequences separately. Finally, the predicted values of both subsequences are recombined to derive the forecast of the original signal. The proposed hybrid model is applied to perform single-step and multi-step predictions on the time series signals from the reactor coolant pump of a specific NPP, and the prediction accuracy is evaluated using metrics such as root mean square error (RMSE), mean absolute percentage error (MAPE), and mean absolute error (MAE). The results demonstrate that the proposed hybrid model can accurately predict and track the operational status of main coolant pump, and comparisons with baseline models highlight the advantages of the proposed hybrid model in complex signal prediction.
2025, 46(4): 225-230.
doi: 10.13832/j.jnpe.2024.090009
Abstract:
A drilling hole on the tube sheet of steam generator (SG) was missed in the manufacturing process, and it was found that one U-shaped tube failed to pass through when most of the U-shaped tubes were expanded after welding or positioning. After research, the equipment manufacturer implemented the scheme of plugging the processed tube hole at the symmetrical position of the missed tube hole. This scheme needs to plug the primary side and the secondary side of the SG tube sheet respectively. Equipment manufacturers adopted the primary side and secondary side respectively plugging scheme. In view of this nonconformity, this paper mainly discusses and analyzes the secondary side plug design, structural strength, flow-induced vibration, weld quality and other aspects from the perspective of nuclear safety review. At the same time, in order to ensure the safety, the failure analysis of the primary side plug, the safety design of the secondary side plug structure and the demonstration of the in-service inspection scheme of the weld quality are also analyzed. The analysis results show that the current tube plugging scheme is reasonable and feasible, but the follow-up inspection in service should be strengthened to ensure the reliability of tube sheet plugging and ensure the safe and stable operation of SG.
A drilling hole on the tube sheet of steam generator (SG) was missed in the manufacturing process, and it was found that one U-shaped tube failed to pass through when most of the U-shaped tubes were expanded after welding or positioning. After research, the equipment manufacturer implemented the scheme of plugging the processed tube hole at the symmetrical position of the missed tube hole. This scheme needs to plug the primary side and the secondary side of the SG tube sheet respectively. Equipment manufacturers adopted the primary side and secondary side respectively plugging scheme. In view of this nonconformity, this paper mainly discusses and analyzes the secondary side plug design, structural strength, flow-induced vibration, weld quality and other aspects from the perspective of nuclear safety review. At the same time, in order to ensure the safety, the failure analysis of the primary side plug, the safety design of the secondary side plug structure and the demonstration of the in-service inspection scheme of the weld quality are also analyzed. The analysis results show that the current tube plugging scheme is reasonable and feasible, but the follow-up inspection in service should be strengthened to ensure the reliability of tube sheet plugging and ensure the safe and stable operation of SG.
2025, 46(4): 237-244.
doi: 10.13832/j.jnpe.2024.070066
Abstract:
To ensure the safe and reliable operation of reactors under ocean conditions, the long-term prediction accuracy of thermal operation parameters under ocean conditions is improved. Based on the thermal operation data of the one-dimensional simulation model of a small modular PWR IP200 under ocean conditions, this study proposes a prediction model combining sequence to sequence (SEQ2SEQ) and autoregressive integrated moving average (ARIMA). First, ARIMA is used to extract the features of the data, and then SEQ2SEQ is used to predict the oscillation value. When the reactor is operating under ocean conditions, it is easy to cause the sloshing of the liquid level inside the system, which in turn causes oscillation in other operating parameters. The prediction results of three thermal operation parameters with different oscillation characteristics, namely, pressurizer pressure, coolant flow, and steam generator steam outlet flow, show that the prediction accuracy is improved by about one order of magnitude, compared with that of using ARIMA model, SEQ2SEQ model and traditional Long Short-Term Memory (LSTM) model alone. The prediction model combining ARIMA and SEQ2SEQ proposed in this study has features of fast calculation speed and high prediction accuracy, which provides an effective method for the prediction of potential failures of small modular PWR under ocean condition.
To ensure the safe and reliable operation of reactors under ocean conditions, the long-term prediction accuracy of thermal operation parameters under ocean conditions is improved. Based on the thermal operation data of the one-dimensional simulation model of a small modular PWR IP200 under ocean conditions, this study proposes a prediction model combining sequence to sequence (SEQ2SEQ) and autoregressive integrated moving average (ARIMA). First, ARIMA is used to extract the features of the data, and then SEQ2SEQ is used to predict the oscillation value. When the reactor is operating under ocean conditions, it is easy to cause the sloshing of the liquid level inside the system, which in turn causes oscillation in other operating parameters. The prediction results of three thermal operation parameters with different oscillation characteristics, namely, pressurizer pressure, coolant flow, and steam generator steam outlet flow, show that the prediction accuracy is improved by about one order of magnitude, compared with that of using ARIMA model, SEQ2SEQ model and traditional Long Short-Term Memory (LSTM) model alone. The prediction model combining ARIMA and SEQ2SEQ proposed in this study has features of fast calculation speed and high prediction accuracy, which provides an effective method for the prediction of potential failures of small modular PWR under ocean condition.
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。
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。
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。
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。
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。
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:
