2024, 45(5): 1-6.
doi: 10.13832/j.jnpe.2024.05.0001
Abstract:
The pebble bed high-temperature gas-cooled reactor (PB-HTGR) is characterized by continuous online refueling. The fuel ball flows slowly in the core of the reactor. The VSOP code, which is widely used in engineering design, employs an approximate and average refueling method to simulate the process, tending to diminish the randomness of the pebble flow of PBR to some extent. In this paper, a new approach of random refueling based on VSOP code is proposed, and it improves the core refueling model and focuses on the impact of the average merging effect of discharging fuel. The results show that the random refueling approach can provide a more refined discharge fuel burnup probability distribution, and the average merging effect of discharging fuel tends to the broadening and overlapping of burnup peaks.
The pebble bed high-temperature gas-cooled reactor (PB-HTGR) is characterized by continuous online refueling. The fuel ball flows slowly in the core of the reactor. The VSOP code, which is widely used in engineering design, employs an approximate and average refueling method to simulate the process, tending to diminish the randomness of the pebble flow of PBR to some extent. In this paper, a new approach of random refueling based on VSOP code is proposed, and it improves the core refueling model and focuses on the impact of the average merging effect of discharging fuel. The results show that the random refueling approach can provide a more refined discharge fuel burnup probability distribution, and the average merging effect of discharging fuel tends to the broadening and overlapping of burnup peaks.
2024, 45(5): 85-91.
doi: 10.13832/j.jnpe.2024.05.0085
Abstract:
Spiral-tube steam generator has the advantages of compact structure and strong heat exchange capacity, and it becomes increasingly prevalent in reactor design. However, its flow and heat transfer characteristics are different from those of straight-tube heat exchanges under marine conditions. Particularly, the instability of two-phase flow under rolling condition remains inadequately explored. In this study, the experimental study on the two-phase flow instability of a single helical tube is carried out under static and rolling conditions, and the process of its transition from single-phase flow to density wave pulsation and then to pressure drop pulsation under different heating power levels is studied. Under the static condition, when the heating power is low, the fluctuation range of each parameter of single-phase flow in the helical tube is within 1%. When the heating power reaches 11 kW, the density wave pulsation with a period of 4.4s is generated, and when the heating power reaches 13 kW, the pressure drop pulsation with a period of about 34.3s is generated. Under the rolling condition, the rolling motion and pulsation have a significant compound effect, and the fluctuation period and amplitude have changed. By studying and processing the experimental data, the characteristics of the period and frequency of the two-phase flow instability in the helical tube are obtained, and the mechanism that causes the difference between the two-phase flow instability in the helical tube and the straight tube flow channel is revealed, as well as the influence mechanism of the rolling condition on the two-phase flow instability.
Spiral-tube steam generator has the advantages of compact structure and strong heat exchange capacity, and it becomes increasingly prevalent in reactor design. However, its flow and heat transfer characteristics are different from those of straight-tube heat exchanges under marine conditions. Particularly, the instability of two-phase flow under rolling condition remains inadequately explored. In this study, the experimental study on the two-phase flow instability of a single helical tube is carried out under static and rolling conditions, and the process of its transition from single-phase flow to density wave pulsation and then to pressure drop pulsation under different heating power levels is studied. Under the static condition, when the heating power is low, the fluctuation range of each parameter of single-phase flow in the helical tube is within 1%. When the heating power reaches 11 kW, the density wave pulsation with a period of 4.4s is generated, and when the heating power reaches 13 kW, the pressure drop pulsation with a period of about 34.3s is generated. Under the rolling condition, the rolling motion and pulsation have a significant compound effect, and the fluctuation period and amplitude have changed. By studying and processing the experimental data, the characteristics of the period and frequency of the two-phase flow instability in the helical tube are obtained, and the mechanism that causes the difference between the two-phase flow instability in the helical tube and the straight tube flow channel is revealed, as well as the influence mechanism of the rolling condition on the two-phase flow instability.
2024, 45(5): 142-154.
doi: 10.13832/j.jnpe.2024.05.0142
Abstract:
To further study the fretting wear of of claddings with the change of oxidation time in practical service, a variety of pre-oxidized claddings were prepared by superheated steam oxidation. In this study, a self-made high-temperature and high-pressure tangential fretting wear tester was used to carry out fretting wear tests simulating the operation conditions of PWR, and the volume wear coefficients of the substrate and the cladding after pre-oxidation at different times were measured. The results show that the surface hardness of the cladding is 2~3 times higher than that of the substrate, and the wear coefficient is reduced by about 90%. A dense oxide layer formed on the surface layer of cladding is an important reason for the change in its wear coefficient. The longer the oxidation time, the thicker the oxide layer, and the cladding with an oxidation time of 200 d has the lowest wear coefficient. In addition, the existence of oxide layer causes the fretting wear mechanism of zircaloy cladding to change from serious abrasive wear and layering to slight abrasive wear and adhesive wear in high temperature and high pressure water environment.
To further study the fretting wear of of claddings with the change of oxidation time in practical service, a variety of pre-oxidized claddings were prepared by superheated steam oxidation. In this study, a self-made high-temperature and high-pressure tangential fretting wear tester was used to carry out fretting wear tests simulating the operation conditions of PWR, and the volume wear coefficients of the substrate and the cladding after pre-oxidation at different times were measured. The results show that the surface hardness of the cladding is 2~3 times higher than that of the substrate, and the wear coefficient is reduced by about 90%. A dense oxide layer formed on the surface layer of cladding is an important reason for the change in its wear coefficient. The longer the oxidation time, the thicker the oxide layer, and the cladding with an oxidation time of 200 d has the lowest wear coefficient. In addition, the existence of oxide layer causes the fretting wear mechanism of zircaloy cladding to change from serious abrasive wear and layering to slight abrasive wear and adhesive wear in high temperature and high pressure water environment.
2024, 45(5): 171-176.
doi: 10.13832/j.jnpe.2024.05.0171
Abstract:
To deeply investigate the intrinsic characteristics of added mass and fluid damping, analyze the impact of viscosity and amplitude on them, and provide guidance for analyzing fluid-induced vibration, this study takes a concentric cylinder as an example to establish a prediction method of added mass and fluid damping based on computational fluid dynamics (CFD). The user-defined function (UDF) is used to set the motion equation of the inner circle, and the overset grid technology is used to complete the grid motion, so as to realize the numerical simulation of the flow field. The shape of the function is determined according to Bearman's hypothesis, and the calculated fluid force curve is fitted by the least square method to obtain the added mass and fluid damping. Finally, the influences of viscosity and dimensionless amplitude on results are compared. The calculation and analysis results show that the viscosity not only affects the fluid damping but also the added mass. The dimensionless amplitude has little effect on the added mass and an obvious effect on the fluid damping. Pressure damping and viscous damping increase in equal proportion with the increase of dimensionless amplitude, and the proportion of pressure damping increases with the decrease of diameter ratio. The solution of the modified formula with dimensionless amplitude effect is in good agreement with the numerical results. The research in this paper has an important guiding role in optimizing the existing analysis methods of flow-induced vibration.
To deeply investigate the intrinsic characteristics of added mass and fluid damping, analyze the impact of viscosity and amplitude on them, and provide guidance for analyzing fluid-induced vibration, this study takes a concentric cylinder as an example to establish a prediction method of added mass and fluid damping based on computational fluid dynamics (CFD). The user-defined function (UDF) is used to set the motion equation of the inner circle, and the overset grid technology is used to complete the grid motion, so as to realize the numerical simulation of the flow field. The shape of the function is determined according to Bearman's hypothesis, and the calculated fluid force curve is fitted by the least square method to obtain the added mass and fluid damping. Finally, the influences of viscosity and dimensionless amplitude on results are compared. The calculation and analysis results show that the viscosity not only affects the fluid damping but also the added mass. The dimensionless amplitude has little effect on the added mass and an obvious effect on the fluid damping. Pressure damping and viscous damping increase in equal proportion with the increase of dimensionless amplitude, and the proportion of pressure damping increases with the decrease of diameter ratio. The solution of the modified formula with dimensionless amplitude effect is in good agreement with the numerical results. The research in this paper has an important guiding role in optimizing the existing analysis methods of flow-induced vibration.
2024, 45(5): 213-218.
doi: 10.13832/j.jnpe.2024.05.0213
Abstract:
Current standards such as ANSI/ANS56.8 and NB/T20018—2021 require the assumption of containment insulation for measuring the leakage rate of containment. The gas quality stability criteria in the standards will fail in the measurement of the leakage rate of non-adiabatic containment. In order to explore the sealing evaluation method of non-adiabatic containment, this study innovatively proposed a leakage rate stability criterion and completed experimental verification on the non-adiabatic scientific research containment at Langfang R&D Base of China Nuclear Power Engineering Co., Ltd. The results indicate that the new stability criterion for leakage rate can be used to measure the leakage rate of non-adiabatic containment. In addition, the test data of a certain nuclear power plant indicates that the new criterion can also be used for measuring the leakage rate of adiabatic containment, shortening the test time. The conclusion of this study can support the application of scientific research non-adiabatic containment to conduct leakage rate research and optimize the sealing test technology of nuclear power plant containment.
Current standards such as ANSI/ANS56.8 and NB/T20018—2021 require the assumption of containment insulation for measuring the leakage rate of containment. The gas quality stability criteria in the standards will fail in the measurement of the leakage rate of non-adiabatic containment. In order to explore the sealing evaluation method of non-adiabatic containment, this study innovatively proposed a leakage rate stability criterion and completed experimental verification on the non-adiabatic scientific research containment at Langfang R&D Base of China Nuclear Power Engineering Co., Ltd. The results indicate that the new stability criterion for leakage rate can be used to measure the leakage rate of non-adiabatic containment. In addition, the test data of a certain nuclear power plant indicates that the new criterion can also be used for measuring the leakage rate of adiabatic containment, shortening the test time. The conclusion of this study can support the application of scientific research non-adiabatic containment to conduct leakage rate research and optimize the sealing test technology of nuclear power plant containment.
2024, 45(5): 237-242.
doi: 10.13832/j.jnpe.2024.05.0237
Abstract:
Based on the present situation of nuclear power measurement system of Qinshan Nuclear Power Phase Ⅱ Units 1&2, focusing on the characteristics and existing problems of the original system, this paper analyzes the necessity of the modification of nuclear power measurement system, and introduces the scope of digital modification of nuclear power measurement system. Through the upgrading design of the nuclear power measurement system of Qinshan Nuclear Power Phase Ⅱ Units 1&2, the design concept, design principle and design flow of the digital modification of the nuclear power measurement system are discussed, and the framework structure design, design characteristics and specific optimization measures of the upgrading of the nuclear power measurement system are given. There were no design changes during the upgrading of the ex-core nuclear measurement system, and the equipment on site was successfully commissioned at the first try and successfully put into operation. The modification scheme and experience can be used as a reference for the change of nuclear measurement system in other nuclear power plants.
Based on the present situation of nuclear power measurement system of Qinshan Nuclear Power Phase Ⅱ Units 1&2, focusing on the characteristics and existing problems of the original system, this paper analyzes the necessity of the modification of nuclear power measurement system, and introduces the scope of digital modification of nuclear power measurement system. Through the upgrading design of the nuclear power measurement system of Qinshan Nuclear Power Phase Ⅱ Units 1&2, the design concept, design principle and design flow of the digital modification of the nuclear power measurement system are discussed, and the framework structure design, design characteristics and specific optimization measures of the upgrading of the nuclear power measurement system are given. There were no design changes during the upgrading of the ex-core nuclear measurement system, and the equipment on site was successfully commissioned at the first try and successfully put into operation. The modification scheme and experience can be used as a reference for the change of nuclear measurement system in other nuclear power plants.
2024, 45(5): 256-261.
doi: 10.13832/j.jnpe.2024.05.0256
Abstract:
When the subchannel code is used for core thermal hydraulic analysis, it is necessary to give the real-time information such as the state parameters of the core inlet, and the computational fluid dynamics (CFD) code can calculate the fine thermal hydraulic parameters of the core inlet. In this paper, the subchannel code CORTH is explicitly coupled with the CFD code FLUENT by means of internal coupling method and dynamic link library technology. The coupling code is further used to simulate the experimental conditions of PNL 2×6 benchmark problem, in which FLUENT calculates the inlet section to provide accurate inlet flow distribution for CORTH, while CORTH calculates the simulated heating section. The results show that the multi-scale coupling code can realize the real-time transmission of thermal parameter information between sub-channel code and CFD code, and the simulation results are in good agreement with the experimental results.
When the subchannel code is used for core thermal hydraulic analysis, it is necessary to give the real-time information such as the state parameters of the core inlet, and the computational fluid dynamics (CFD) code can calculate the fine thermal hydraulic parameters of the core inlet. In this paper, the subchannel code CORTH is explicitly coupled with the CFD code FLUENT by means of internal coupling method and dynamic link library technology. The coupling code is further used to simulate the experimental conditions of PNL 2×6 benchmark problem, in which FLUENT calculates the inlet section to provide accurate inlet flow distribution for CORTH, while CORTH calculates the simulated heating section. The results show that the multi-scale coupling code can realize the real-time transmission of thermal parameter information between sub-channel code and CFD code, and the simulation results are in good agreement with the experimental results.
2018, 39(5): 1-3.
doi: 10.13832/j.jnpe.2018.05.0001
摘要：
主要介绍了我国在建、在运核电机组的基本状况和最新进展，以及我国在提升核设施安全水平方面的相关措施。在国家能源局印发的《能源技术创新“十三五”规划》要求之下，我国推出一系列先进核能和小型堆的发展计划，开展了“海洋核动力平台示范工程建设”并建立相关标准。最后总结了中国核电目前面临的挑战和未来的展望。
主要介绍了我国在建、在运核电机组的基本状况和最新进展，以及我国在提升核设施安全水平方面的相关措施。在国家能源局印发的《能源技术创新“十三五”规划》要求之下，我国推出一系列先进核能和小型堆的发展计划，开展了“海洋核动力平台示范工程建设”并建立相关标准。最后总结了中国核电目前面临的挑战和未来的展望。
2019, 40(4): 1-8.
摘要：
热管冷却反应堆采用固态反应堆设计理念，通过热管非能动方式导出堆芯热量。本文总结了热管冷却反应堆的概念初创、积极探索、重大突破的发展历程；分析了热管冷却反应堆的技术特点，包括固态属性、固有安全性高、运行特性简单、易于模块化与易扩展和运输特性良好等核心优势；归纳了热管冷却反应堆中热管性能、材料工艺、能量转换等技术现状，并提出热管冷却反应堆进一步发展将面临的材料、制造工艺、运行可维护性等挑战，从而明确了热管冷却反应堆未来的发展趋势，为革新型热管冷却反应堆技术的发展与应用提供良好的方向指引。总体而言，热管冷却反应堆在深空探测与推进、陆基核电源、深海潜航探索等场景中具有广阔的应用前景，有可能成为改变未来核动力格局的颠覆性技术之一。
热管冷却反应堆采用固态反应堆设计理念，通过热管非能动方式导出堆芯热量。本文总结了热管冷却反应堆的概念初创、积极探索、重大突破的发展历程；分析了热管冷却反应堆的技术特点，包括固态属性、固有安全性高、运行特性简单、易于模块化与易扩展和运输特性良好等核心优势；归纳了热管冷却反应堆中热管性能、材料工艺、能量转换等技术现状，并提出热管冷却反应堆进一步发展将面临的材料、制造工艺、运行可维护性等挑战，从而明确了热管冷却反应堆未来的发展趋势，为革新型热管冷却反应堆技术的发展与应用提供良好的方向指引。总体而言，热管冷却反应堆在深空探测与推进、陆基核电源、深海潜航探索等场景中具有广阔的应用前景，有可能成为改变未来核动力格局的颠覆性技术之一。
2019, 40(6): 1-6.
doi: 10.13832/j.jnpe.2019.06.0001
摘要：
放射性废液得到有效处理是世界各国核工业迅猛发展的前提，其关键技术的现状和发展方向也是我国核工业界关注的焦点。本文介绍了几种放射性废液处理的传统方法及涌现出的新技术，概述了各种方法的原理及优、缺点，同时讨论了放射性废液处理技术今后的研究方向及发展趋势。
放射性废液得到有效处理是世界各国核工业迅猛发展的前提，其关键技术的现状和发展方向也是我国核工业界关注的焦点。本文介绍了几种放射性废液处理的传统方法及涌现出的新技术，概述了各种方法的原理及优、缺点，同时讨论了放射性废液处理技术今后的研究方向及发展趋势。
2019, 40(3): 103-108.
doi: 10.13832/j.jnpe.2019.03.0103
摘要：
以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明：四取中逻辑在可靠性方面优化程度相对较高。
以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明：四取中逻辑在可靠性方面优化程度相对较高。
2019, 40(1): 1-7.
doi: 10.13832/j.jnpe.2019.01.0001
摘要：
“华龙一号”是我国自主设计研发的具有完整知识产权的第三代百万千瓦级压水堆核电技术。本文介绍了“华龙一号”的产生历程，系统论述了“华龙一号”反应堆堆芯与安全设计特点，包括“华龙一号”研发过程中开展的堆芯核设计、热工水力设计、安全设计、设计验证及“华龙一号”持续开展的设计改进与优化等内容，通过采用新的设计理念和设计技术，全面提高了“华龙一号”作为三代核电技术的经济性、灵活性和安全性。
“华龙一号”是我国自主设计研发的具有完整知识产权的第三代百万千瓦级压水堆核电技术。本文介绍了“华龙一号”的产生历程，系统论述了“华龙一号”反应堆堆芯与安全设计特点，包括“华龙一号”研发过程中开展的堆芯核设计、热工水力设计、安全设计、设计验证及“华龙一号”持续开展的设计改进与优化等内容，通过采用新的设计理念和设计技术，全面提高了“华龙一号”作为三代核电技术的经济性、灵活性和安全性。
2020, 41(5): 132-137.
摘要：
为解决核电厂传统监测手段的局限性，提出将核主元分析法（KPCA）引入核电厂设备在线监测领域中，并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性，将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明，KPCA算法可适应核电厂设备监测的要求，能比现有阈值监测手段提供更为早期的故障预警。同时，相比于常规的主元分析法（PCA），KPCA算法能够提取各变量之间的非线性关系，识别出设备不同的运行模式，有效减少误报警。
为解决核电厂传统监测手段的局限性，提出将核主元分析法（KPCA）引入核电厂设备在线监测领域中，并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性，将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明，KPCA算法可适应核电厂设备监测的要求，能比现有阈值监测手段提供更为早期的故障预警。同时，相比于常规的主元分析法（PCA），KPCA算法能够提取各变量之间的非线性关系，识别出设备不同的运行模式，有效减少误报警。
2019, 40(1): 74-77.
doi: 10.13832/j.jnpe.2019.01.0074
摘要：
介绍了中广核研究院在事故容错燃料（ATF）包壳领域的最新成果，通过预置粉末式脉冲激光熔覆技术，在不同的功率下制备出不同厚度的锆包壳管Cr保护层；通过高温蒸汽氧化增重数据发现，采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能，证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构，采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
介绍了中广核研究院在事故容错燃料（ATF）包壳领域的最新成果，通过预置粉末式脉冲激光熔覆技术，在不同的功率下制备出不同厚度的锆包壳管Cr保护层；通过高温蒸汽氧化增重数据发现，采用半导体脉冲激光熔覆技术、脉冲激光功率50~60 W、螺距0.8~0.9 mm、角速度10°/s等参数条件下制备Cr涂层可以获得较好的抗高温氧化性能，证明保护的效果直接受涂层质量控制。通过SEM分析了涂层的显微结构，采用扩散机理解释了Cr涂层在1200℃下与锆合金基体相容性良好的原因。
2020, 41(3): 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。
2020, 41(4): 191-196.
摘要：
热管冷却反应堆（简称“热管堆”）高温运行下的结构热膨胀效应会显著影响反应堆的传热和中子物理输运过程。本文提出了一种考虑固体堆芯显著膨胀的几何更新和反应性反馈方法，并构建了基于动态几何的中子物理/热工/力学3场核热力耦合分析程序。在核热力耦合中主要考虑温度引起微观截面的变化、材料密度的变化以及热膨胀引起堆芯尺寸的变化。基于提出的核热力耦合方法，对MegaPower热管堆进行了核热力耦合分析，分析了不同松弛因子下，堆芯功率分布和径向功率因子的收敛性。核热力计算表明，热膨胀造成堆芯边通道的中子泄漏增加，从而产生负反应性反馈；同时，边通道中子泄漏增加加剧了功率分布的不均匀性，传热恶化，考虑核热力耦合后，径向功率因子从非耦合情形的1.20提升到1.23，燃料峰值温度增加11 K。
热管冷却反应堆（简称“热管堆”）高温运行下的结构热膨胀效应会显著影响反应堆的传热和中子物理输运过程。本文提出了一种考虑固体堆芯显著膨胀的几何更新和反应性反馈方法，并构建了基于动态几何的中子物理/热工/力学3场核热力耦合分析程序。在核热力耦合中主要考虑温度引起微观截面的变化、材料密度的变化以及热膨胀引起堆芯尺寸的变化。基于提出的核热力耦合方法，对MegaPower热管堆进行了核热力耦合分析，分析了不同松弛因子下，堆芯功率分布和径向功率因子的收敛性。核热力计算表明，热膨胀造成堆芯边通道的中子泄漏增加，从而产生负反应性反馈；同时，边通道中子泄漏增加加剧了功率分布的不均匀性，传热恶化，考虑核热力耦合后，径向功率因子从非耦合情形的1.20提升到1.23，燃料峰值温度增加11 K。
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:
2018, 39(3): 6-9.
doi: 10.13832/j.jnpe.2018.03.0006
Abstract:
2019, 40(5): 1-5.
Abstract:
2022, 43(2): 1-8.
doi: 10.13832/j.jnpe.2022.02.0001
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