2024, 45(3): 1-13.
doi: 10.13832/j.jnpe.2024.03.0001
Abstract:
Reactor decommissioning is one of the nuclear facility decommissioning work that needs to be focused on. In this paper, the overall situation of global reactor decommissioning is analyzed. The concept of decommissioning technology in a broad sense is put forward, and a reactor decommissioning technology system consisting of decommissioning management and overall technology, decommissioning special technology and common support technology is constructed. The policies, regulations and standards, technical route, project management and other decommissioning management and overall technology are discussed. The special decommissioning technologies such as safe shutdown, characteristic investigation, decontamination, cutting and dismantling, waste management and final decommissioning management are analyzed. The common supporting technologies such as digitization and intellectualization, radiation protection and monitoring are discussed. After comprehensive analysis and demonstration, the development direction of specific decommissioning technology and the development trend of reactor decommissioning technology in the future are prospected.
Reactor decommissioning is one of the nuclear facility decommissioning work that needs to be focused on. In this paper, the overall situation of global reactor decommissioning is analyzed. The concept of decommissioning technology in a broad sense is put forward, and a reactor decommissioning technology system consisting of decommissioning management and overall technology, decommissioning special technology and common support technology is constructed. The policies, regulations and standards, technical route, project management and other decommissioning management and overall technology are discussed. The special decommissioning technologies such as safe shutdown, characteristic investigation, decontamination, cutting and dismantling, waste management and final decommissioning management are analyzed. The common supporting technologies such as digitization and intellectualization, radiation protection and monitoring are discussed. After comprehensive analysis and demonstration, the development direction of specific decommissioning technology and the development trend of reactor decommissioning technology in the future are prospected.
2024, 45(3): 14-19.
doi: 10.13832/j.jnpe.2024.03.0014
Abstract:
Nuclear reactors involve a series of multi-physics coupling processes with complex interactions. With the rapid development of high-performance computing technology, the analysis based on multi-physics coupling has been paid more and more attention. Based on the multi-physics coupling platform MOOSE, the steady-state and transient multi-physics coupling simulation of Xi'an pulsed reactor is studied. The nonlinear problems in multi-physics coupling are solved by Picard iteration and Jacobian Free Newton-Krylov (JFNK) method, and the multi-physical coupling calculation of three-dimensional neutron spatio-temporal dynamics, three-dimensional solid heat conduction and one-dimensional fluid flow and heat transfer is realized. The reactor behavior under 2 MW steady-state operation and 3.45$ \$$ (1$\$$ represents an effective delayed neutron fraction) pulse operation of Xi’an pulsed reactor is calculated, and the three-dimensional power and temperature distribution of the core are obtained. The calculated results are in good agreement with the experimental results, which proves the correctness of multi-physics coupling. The multi-physics coupling method developed in this paper has the advantages of good geometric adaptability and flexible coupling, and has the potential to be applied to other micro reactors.
Nuclear reactors involve a series of multi-physics coupling processes with complex interactions. With the rapid development of high-performance computing technology, the analysis based on multi-physics coupling has been paid more and more attention. Based on the multi-physics coupling platform MOOSE, the steady-state and transient multi-physics coupling simulation of Xi'an pulsed reactor is studied. The nonlinear problems in multi-physics coupling are solved by Picard iteration and Jacobian Free Newton-Krylov (JFNK) method, and the multi-physical coupling calculation of three-dimensional neutron spatio-temporal dynamics, three-dimensional solid heat conduction and one-dimensional fluid flow and heat transfer is realized. The reactor behavior under 2 MW steady-state operation and 3.45$ \$$ (1$\$$ represents an effective delayed neutron fraction) pulse operation of Xi’an pulsed reactor is calculated, and the three-dimensional power and temperature distribution of the core are obtained. The calculated results are in good agreement with the experimental results, which proves the correctness of multi-physics coupling. The multi-physics coupling method developed in this paper has the advantages of good geometric adaptability and flexible coupling, and has the potential to be applied to other micro reactors.
2024, 45(3): 139-145.
doi: 10.13832/j.jnpe.2024.03.0139
Abstract:
In order to predict the steam oxidation behavior of FeCrAl alloy at different temperatures and provide the model for the evolution simulation of the performance of FeCrAl cladding under loss of coolant accident (LOCA), a two-stage differential oxidation weight-gain model was proposed based on the reaction and diffusion control mechanisms, and a parameter calibration method was also presented. Combined with the experimental data from FeCrAl steam oxidation tests at high temperature (900-1200℃) and medium temperature (400℃), the model can uniformly describe the weight-gain behavior of FeCrAl alloy in the temperature range of 400-1200℃, and the error with experimental data is controlled within 20%. At the same time, it is observed that the critical weight-gain of the reaction-diffusion mechanism is basically unchanged at 400-900℃, but increases significantly at higher temperature, because the oxidation layer grows too fast to form the dense oxidation protective layer. In addition, considering the influence of initial oxide layer from water corrosion and the change of steam pressure during LOCA, a modified scheme of the oxidation-weight gain model is given. This study is expected to provide oxidation model and parameters for the failure behavior simulation of the FeCrAl alloy cladding under LOCA accidents.
In order to predict the steam oxidation behavior of FeCrAl alloy at different temperatures and provide the model for the evolution simulation of the performance of FeCrAl cladding under loss of coolant accident (LOCA), a two-stage differential oxidation weight-gain model was proposed based on the reaction and diffusion control mechanisms, and a parameter calibration method was also presented. Combined with the experimental data from FeCrAl steam oxidation tests at high temperature (900-1200℃) and medium temperature (400℃), the model can uniformly describe the weight-gain behavior of FeCrAl alloy in the temperature range of 400-1200℃, and the error with experimental data is controlled within 20%. At the same time, it is observed that the critical weight-gain of the reaction-diffusion mechanism is basically unchanged at 400-900℃, but increases significantly at higher temperature, because the oxidation layer grows too fast to form the dense oxidation protective layer. In addition, considering the influence of initial oxide layer from water corrosion and the change of steam pressure during LOCA, a modified scheme of the oxidation-weight gain model is given. This study is expected to provide oxidation model and parameters for the failure behavior simulation of the FeCrAl alloy cladding under LOCA accidents.
2024, 45(3): 154-160.
doi: 10.13832/j.jnpe.2024.03.0154
Abstract:
A large-load isolator is designed for heavy equipment of nuclear power system by combining disc spring with magnetorheological liquid phase. The magnetorheological fluid damper is used as the main damping component, and the composite component of small stiffness and large stiffness disc spring is used as the main bearing structure. Through the analysis and design of the disc spring composite component, the structural parameters of the composite component are determined, and the sample isolator is formed accordingly. The spring rate ratio and damping ratio of the isolator samples were tested. The results show that the rated load of the isolator reaches 11 tons and the natural frequency is about 6.2 Hz. The dynamic stiffness and damping ratio performance of the isolator are relatively stable in the range of main vibration isolation frequencies. However, the dynamic stiffness and damping ratio are greatly affected by the amplitude. When the vibration amplitude increases from 0.05 mm to 0.2 mm, the dynamic stiffness decreases from about 100 kN/mm to about 45 kN/mm, and the damping ratio increases from about 0.07 to about 0.19. The research in this paper can provide technical reference and support for subsequent engineering applications.
A large-load isolator is designed for heavy equipment of nuclear power system by combining disc spring with magnetorheological liquid phase. The magnetorheological fluid damper is used as the main damping component, and the composite component of small stiffness and large stiffness disc spring is used as the main bearing structure. Through the analysis and design of the disc spring composite component, the structural parameters of the composite component are determined, and the sample isolator is formed accordingly. The spring rate ratio and damping ratio of the isolator samples were tested. The results show that the rated load of the isolator reaches 11 tons and the natural frequency is about 6.2 Hz. The dynamic stiffness and damping ratio performance of the isolator are relatively stable in the range of main vibration isolation frequencies. However, the dynamic stiffness and damping ratio are greatly affected by the amplitude. When the vibration amplitude increases from 0.05 mm to 0.2 mm, the dynamic stiffness decreases from about 100 kN/mm to about 45 kN/mm, and the damping ratio increases from about 0.07 to about 0.19. The research in this paper can provide technical reference and support for subsequent engineering applications.
2024, 45(3): 206-212.
doi: 10.13832/j.jnpe.2024.03.0206
Abstract:
Aiming at the problems of blade sticking and damper body strength failure of sodium to air heat exchanger in nuclear power sodium-cooled fast reactor under high temperature working environment, a double-blade damper for sodium to air heat exchanger of sodium-cooled fast reactor was designed. The characteristics of flow field, temperature distribution of damper body, deformation and stress of damper body with different opening degrees were studied using fluid-thermal-solid coupling method. The results show that when the opening of damper is below 45 degrees, the fluid will have obvious velocity gradient and pressure gradient before and after it passes through damper. The greater the local temperature difference, the greater the thermal stress value, and the maximum stress is 206.94 MPa, which appears at the edge of the frame side sealing plate, and the maximum stress value meets the material strength requirements. The deformation and stress of damper body are mainly thermal deformation and thermal stress caused by heating. The maximum deformation of damper is 3.3368 mm, and the deformation of the frame is greater than that of the blade in all directions. The newly designed damper has no phenomenon of blade sticking.
Aiming at the problems of blade sticking and damper body strength failure of sodium to air heat exchanger in nuclear power sodium-cooled fast reactor under high temperature working environment, a double-blade damper for sodium to air heat exchanger of sodium-cooled fast reactor was designed. The characteristics of flow field, temperature distribution of damper body, deformation and stress of damper body with different opening degrees were studied using fluid-thermal-solid coupling method. The results show that when the opening of damper is below 45 degrees, the fluid will have obvious velocity gradient and pressure gradient before and after it passes through damper. The greater the local temperature difference, the greater the thermal stress value, and the maximum stress is 206.94 MPa, which appears at the edge of the frame side sealing plate, and the maximum stress value meets the material strength requirements. The deformation and stress of damper body are mainly thermal deformation and thermal stress caused by heating. The maximum deformation of damper is 3.3368 mm, and the deformation of the frame is greater than that of the blade in all directions. The newly designed damper has no phenomenon of blade sticking.
2024, 45(3): 246-251.
doi: 10.13832/j.jnpe.2024.03.0246
Abstract:
Liquid lead-bismuth alloy has the characteristics of good thermal conductivity and high heat capacity, making it an ideal coolant for the new generation of advanced reactor. In this article, a full-scale computational fluid dynamics (CFD) model for plate fuel assembly in high-flow lead-bismuth environment was established, the transient fluid dynamics analysis based on large eddy simulation (LES) turbulence model was carried out and the fluid excitation force on the fuel plate was obtained. The dynamic analysis model of fuel plate was established, the structural dynamics calculation based on time domain was carried out according to the transient fluid excitation data, and the displacement response of fuel plate was obtained. The calculation results show that the fluid excitation force on the fuel plate in the middle position is much greater than that on the two sides because of the vortex shedding formed by the hoisting structure. The displacement response of the fuel plate concentrates on its own first-order frequency, and the first-order frequency of single fuel plate is much greater than the main frequency of turbulent excitation, so there is no risk of resonance of the fuel plate under fluid excitation. Considering the influence of inlet turbulence intensity, the conservatism of the flow induced vibration analysis method based on rectangular channel power density spectrum may be insufficient. This research can provide a reference for the development of new generation high-performance fuel assemblies.
Liquid lead-bismuth alloy has the characteristics of good thermal conductivity and high heat capacity, making it an ideal coolant for the new generation of advanced reactor. In this article, a full-scale computational fluid dynamics (CFD) model for plate fuel assembly in high-flow lead-bismuth environment was established, the transient fluid dynamics analysis based on large eddy simulation (LES) turbulence model was carried out and the fluid excitation force on the fuel plate was obtained. The dynamic analysis model of fuel plate was established, the structural dynamics calculation based on time domain was carried out according to the transient fluid excitation data, and the displacement response of fuel plate was obtained. The calculation results show that the fluid excitation force on the fuel plate in the middle position is much greater than that on the two sides because of the vortex shedding formed by the hoisting structure. The displacement response of the fuel plate concentrates on its own first-order frequency, and the first-order frequency of single fuel plate is much greater than the main frequency of turbulent excitation, so there is no risk of resonance of the fuel plate under fluid excitation. Considering the influence of inlet turbulence intensity, the conservatism of the flow induced vibration analysis method based on rectangular channel power density spectrum may be insufficient. This research can provide a reference for the development of new generation high-performance fuel assemblies.
2024, 45(3): 263-267.
doi: 10.13832/j.jnpe.2024.03.0263
Abstract:
Aiming at the phenomenon of over-range triggered quality bit anomaly of all three loop flowmeters during single pump operation in the commissioning of Hainan 1&2 units, the measurement method of loop flow rate and the measured data of flow signal in cold shutdown state are analyzed and studied based on the measurement principle of reactor coolant flow signal, shutdown protection logic and quality bit setting principle. This paper points out that the physical range of flow signal currently designed cannot envelope the relative flow rate value under various conditions. According to the analysis, it is necessary to adjust the measurement range of the flow meter so that the corresponding process flow range after converting the output 4~20 mA current is adjusted from 0~120%FP (FP is full power) to 0~129%FP (the loop flow signal shows X%FP, indicating that the current flow is X% of the relative flow rate during full power operation), and the current output of the flowmeters during the normal full power operation should be calibrated to 13.615 mA.
Aiming at the phenomenon of over-range triggered quality bit anomaly of all three loop flowmeters during single pump operation in the commissioning of Hainan 1&2 units, the measurement method of loop flow rate and the measured data of flow signal in cold shutdown state are analyzed and studied based on the measurement principle of reactor coolant flow signal, shutdown protection logic and quality bit setting principle. This paper points out that the physical range of flow signal currently designed cannot envelope the relative flow rate value under various conditions. According to the analysis, it is necessary to adjust the measurement range of the flow meter so that the corresponding process flow range after converting the output 4~20 mA current is adjusted from 0~120%FP (FP is full power) to 0~129%FP (the loop flow signal shows X%FP, indicating that the current flow is X% of the relative flow rate during full power operation), and the current output of the flowmeters during the normal full power operation should be calibrated to 13.615 mA.
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。
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。
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。
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。
2019, 40(4): 127-133.
摘要：
为提高核主泵在全工况点的数值模拟精度，研究了数值模拟过程中近壁面网格尺度、湍流模型、流动状态3种因素对计算精度的影响。结果表明，在定常状态下，重整化群（RNG） k-ε湍流模型和标准壁面函数法在近壁面网格尺度（y+）为50左右时具有较高的计算精度，并且其计算精度高于RNG k-ε增强壁面函数法、低雷诺数k-ε和剪切应力传输（SST）k-ω这3种湍流模型的计算精度，但上述不同网格尺度和湍流模型的计算结果均存在较大的计算误差；采用非定常计算时的计算精度明显高于定常计算，能够反映出扬程曲线在关死点附近的驼峰现象，效率的计算精度也有一定改善，更适合于对核主泵进行性能预测。
为提高核主泵在全工况点的数值模拟精度，研究了数值模拟过程中近壁面网格尺度、湍流模型、流动状态3种因素对计算精度的影响。结果表明，在定常状态下，重整化群（RNG） k-ε湍流模型和标准壁面函数法在近壁面网格尺度（y+）为50左右时具有较高的计算精度，并且其计算精度高于RNG k-ε增强壁面函数法、低雷诺数k-ε和剪切应力传输（SST）k-ω这3种湍流模型的计算精度，但上述不同网格尺度和湍流模型的计算结果均存在较大的计算误差；采用非定常计算时的计算精度明显高于定常计算，能够反映出扬程曲线在关死点附近的驼峰现象，效率的计算精度也有一定改善，更适合于对核主泵进行性能预测。
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:
2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
Abstract:
2021, 42(4): 1-5.
doi: 10.13832/j.jnpe.2021.04.0001
Abstract:
2022, 43(4): 1-4.
doi: 10.13832/j.jnpe.2022.04.0001
Abstract:
2019, 40(5): 1-5.
Abstract:
2018, 39(3): 6-9.
doi: 10.13832/j.jnpe.2018.03.0006
Abstract:
2019, 40(4): 1-8.
Abstract:
2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
Abstract:
