分类： 核科学技术 >> 核科学与技术 提交时间： 2023-06-21
摘要： 熔盐堆作为第四代先进反应堆的重要堆型之一，以高沸点熔盐为核燃料，具有高温输出、常压操作等特点。而基于温差发电的热管熔盐堆，兼具了熔盐堆、热管和温差发电的优势，具有输出温度高、热电转换效率高、结构简单及安全可靠等优点。故该堆型在能源系统领域具有极大的优势，是外太空及深海探测任务的理想能源。但因堆芯熔盐低热导率而形成的热管密集排布给热管冷凝段温差发电传热设计带来了难题。针对该堆型设计需求，本文提出适于熔盐堆的热管-温差发电耦合系统结构并进行了传热分析。堆芯热管冷凝段采用塔式温差发电系统结构设计，整体热端座与堆芯热管冷凝端相配合，形成从下至上的第1层至第N层热段套；冷端座套设于热端座外，内设冷端热管通道；热端座的外侧壁与冷端座的内侧壁之间贴有温差发电片，发电片间隙采用保温棉减少漏热。采用Ansys Workbench开展了适于热管熔盐堆的四层塔式温差发电系统传热仿真模拟，分析表明，系统运行时，整体塔座温度分布均匀，热量有效利用率大于 96%，系统漏热量小于4%，发电片两侧温差大于490℃，利于提高热电转换效率，设计具有可行性，有利于推动温差发电在热管熔盐堆中的应用。
分类： 核科学技术 >> 核科学技术其他学科 提交时间： 2025-04-18
摘要： 热管微型熔盐堆适用于深海堆、空间堆等特种应用。为有效验证微型熔盐堆的设计方案、关键技术及系统集成能力，本文搭建了高温热管传热实验装置，深入开展热管启动特性、热管温差发电耦合传热等方面的研究。实验中，实验装置在热态加盐启动。实验结果的分析表明，因辐射传热与自然对流的作用，700℃下熔盐等效热导率达 11.2 W·(m·K)−1。建立了高温热管传热实验装置的热阻网络模型，模型热阻与实验结果误差为 44.9%，说明当前模型存在局限，但熔盐热管热阻与实验结果仅相差19.3%，热阻网络法与实验结合的方法仍具可行性。温差发电系统传热热阻为0.51K/W，占总热阻的87.3%，优化温差发电传热是提高熔盐堆传热效率的关键。
分类： 物理学 >> 电磁学、光学、声学、传热、经典力学和流体动力学 提交时间： 2025-02-11
摘要： 热管微型熔盐堆适用于深海堆、空间堆等特种应用。为有效验证微型熔盐堆的设计方案、关键技术及系统集成能力，本文搭建了高温热管传热实验装置，深入开展热管启动特性、热管温差发电耦合传热等方面的研究。实验中，实验装置在热态加盐启动。实验结果的分析表明，因辐射传热与自然对流的作用，700℃下熔盐等效热导率达 11.2 W/m∙K 。建立了高温热管传热实验装置的热阻网络模型，模型热阻与实验结果误差为 44.9%，说明当前模型存在局限，但熔盐热管热阻与实验结果仅相差19.3%，热阻网络法与实验结合的方法仍具可行性。温差发电系统传热热阻为0.51K/W，占总热阻的87.3%，优化温差发电传热是提高熔盐堆传热效率的关键。
分类： 核科学技术 >> 核科学与技术 提交时间： 2024-08-15
摘要： 熔盐堆是六种第四代反应堆之一，其使用液态核燃料，与常规固态核燃料堆型不一样。为做好堆芯设计管理和核安全监管，有必要对其7Li丰度与核临界参数的关系进行分析。本研究通过参考具有工程实践的熔盐堆进行建模，运用软件仿真模拟计算，分析不同7Li丰度燃料盐对熔盐堆反应性的影响，分析核临界参数变化的规律。通过迭代计算，快速精确找到熔盐堆核临界时的7Li丰度值。最终得出结论，熔盐堆反应性随着燃料盐7Li丰度的增加而增加，且在熔盐堆反应性变化率也与7Li丰度有关。根据本研究分析成果进行深入探讨，从法律法规角度总结相关管理要求，从审管角度提出应该关注的建议要点。
分类： 核科学技术 >> 核科学与技术 提交时间： 2024-07-15
摘要： 熔盐堆是六种第四代反应堆之一，其使用液态核燃料，与常规固态核燃料堆型不一样。为做好堆芯设计管理和核安全监管，有必要对其7Li丰度与核临界参数的关系进行分析。本研究通过参考具有工程实践的熔盐堆进行建模，运用软件仿真模拟计算，分析不同7Li丰度燃料盐对熔盐堆反应性的影响，分析核临界参数变化的规律。通过迭代计算，快速精确找到熔盐堆核临界时的7Li丰度值。最终得出结论，熔盐堆反应性随着燃料盐7Li丰度的增加而增加，且在熔盐堆反应性变化率也与7Li丰度有关。根据本研究分析成果进行深入探讨，从法律法规角度总结相关管理要求，从审管角度提出应该关注的建议要点。
分类： 物理学 >> 核物理学 提交时间： 2025-04-16
摘要： The 2019 edition of the International Reactor Physics Evaluation Project (IRPhEP) Handbook incorporated the Molten Salt Reactor Experiment (MSRE) benchmark, providing keff (effective multiplication factor) values derived from first criticality experiments and control rod worth calculations for multiple nuclear data libraries including ENDF/B-VII.1. This benchmark constitutes the first comprehensive reference case for molten salt reactor physics, having been extensively utilized to assess the consistency and accuracy of Monte Carlo codes and nuclear data libraries in molten salt reactor modeling. Since 2011, the Thorium Molten Salt Reactor (TMSR) nuclear energy system has been under development at the Shanghai Institute of Applied Physics, Chinese Academy of Sciences to facilitate thorium resource utilization. In support of this initiative, the China Nuclear Data Center developed specialized CENDL-TMSR-V1 libraries tailored for thorium-uranium fuel cycles. Nevertheless, the verification status of Chinese nuclear libraries CENDL-3.2 and CENDL-TMSR-V1 in molten salt reactor applications remains unexplored. In this work, a high-fidelity MSRE model was developed using OpenMC, with comparative analyses conducted across four evaluated nuclear data libraries: ENDF/B-VII.1, ENDF/B-VIII.0, CENDL-3.2, and CENDL-TMSR-V1. A systematic evaluation of neutronic parameters was performed, encompassing reactivity coefficients, control rod differential worth, zero-power flux distribution, and 500-day burn-up calculations. Key findings reveal that: The relative deviations in keff between all libraries and IRPhEP benchmark values remain below 300 pcm (0.3% Δk/k). The maximum relative discrepancy in power distribution predictions between CENDL-series libraries and ENDF/B-VII.1 is <2%. The keff deviations during burn-up calculations are maintained within 0.2%/. This study validates the applicability of CENDL-series libraries for molten salt reactor neutronic simulations.
分类： 核科学技术 >> 核科学与技术 提交时间： 2025-04-01
摘要： 熔盐堆具有固有安全性和部署灵活性，能够广泛应用于多种场景。功率控制系统是确保熔盐堆安全稳定运行的关键系统之一，其在控制堆芯燃料盐温度的同时，还需要对负载变化进行快速响应，以满足基于熔盐堆的多样化复合能源系统中的潜在需求。控制目标之间存在相互影响，且执行器的数量和可调范围往往有限，因此给控制器参数的选择带来了巨大挑战。本文使用反应堆功率控制系统上普遍采用的比例-积分-微分(Proportional-Integral-Derivative, PID)控制器对堆芯温度和功率进行控制。在MATLAB/Simulink软件中建立熔盐增殖堆(Molten Salt Breeder Reactor, MSBR)一回路非线性模型。基于该模型采用多目标粒子群算法(Multi-Objective Particle Swarm Optimization, MOPSO)对控制器参数进行了优化，并根据灵敏度比偏差筛选每个目标都可接受的优解，实现了自动化选取控制器参数的最佳方案。通过仿真不同工况对控制器的性能进行了验证，结果表明，多目标粒子群算法优化后的控制器具有良好的负荷跟踪能力，能够实现功率及温度双目标的平衡，在负载阶跃变化时能抑制超调量在2%范围内，响应速度快，抗干扰能力强。
分类： 核科学技术 >> 核科学与技术 提交时间： 2025-04-29
摘要： 液体燃料熔盐堆是第四代先进核能系统的候选堆型之一，应急排盐冷却系统是其可设置的特殊安全设计。本文以熔盐实验堆（Molten Salt Reactor Experiment, MSRE）应急排盐冷却系统为研究对象，探究换热元件内层屏障破损对其正常余热导出功能的影响。该研究基于计算流体动力学模拟方法，通过构建传热和多相流动模型分析破损后的传热和流动现象，并对关键影响参数进行敏感性分析。结果表明，内层套管破损后，18.4%的冷却水从破损位置经气隙层流出，单根换热元件功率提升到29.434 kW，破损位置附近外层套管出现局部温度低点。敏感性分析结果表明，破损尺寸、破损位置和气隙层压力变化均会对换热元件传热产生明显影响，其中传热对破损尺寸敏感性最强。该分析结果可以为熔盐堆应急排盐冷却系统工程设计提供研究数据。
分类： 物理学 >> 核物理学 提交时间： 2025-05-14
Cao, Mr. Jintong
Zhu, Dr. Guifeng
Yu, Mr. Changqing
Liu, Dr. Yafen
Zou, Dr. Yang 邹杨
Yan, Dr. Rui
Xu, Prof. Hongjie
摘要： Molten salt reactors, being the only reactor type among Generation IV advanced nuclear reactors to utilize liquid fuel, offer inherent safety, high-temperature and low-pressure operation, as well as the capability for online fuel reprocessing. However, fuel salt flow results in the decay of delayed neutron precursors (DNPs) outside the core, leading to fluctuations in the effective delayed neutron fraction and consequently impacting reactor reactivity. Particularly under accident scenarios—such as combined pump shutdown and inability to rapidly scram the reactor—the reliance solely on negative temperature feedback may cause a substantial increase in core temperature, posing a threat to reactor safety. To address these issues, this paper introduces an innovative design for a passive fluid-driven Suspended Control Rod (SCR) aimed at dynamically compensating for reactivity fluctuations caused by DNPs flowing with fuel flow. The control rod operates passively by leveraging the combined effects of gravity, buoyancy, and fluid dynamic forces, thereby eliminating the need for any external drive mechanism and allowing direct integration within the core’s active region. Using a 150 MWth thorium-based molten salt reactor as the reference design, a mathematical model was developed to systematically analyze the effects of key parameters—including the SCR's geometric dimensions and density—on its performance, examine its motion characteristics under different core flow conditions, and assess its feasibility for dynamic compensation of reactivity changes caused by fuel flow. The study’s results demonstrate that the SCR can effectively counteract the reactivity fluctuations induced by fuel flow within molten salt reactors. Sensitivity analysis revealed that the SCR’s average density exerts a profound impact on its start-up flow threshold, channel flow rate, resistance to fuel density fluctuations, and response characteristics, underscoring the critical need to optimize this parameter. Moreover, by judiciously selecting the SCR’s length, number of deployed units, and placement, one can achieve the necessary reactivity control while also maintaining a favorable balance between neutron economy and heat transfer performance. Ultimately, this study provides an innovative solution for passive reactivity control in molten salt reactors, offering substantial potential for practical engineering applications.
分类： 物理学 >> 核物理学 提交时间： 2025-03-31
Cao, Mr. Jintong
Zhu, Dr. Guifeng
Yu, Mr. Changqing
Liu, Dr. Yafen
Zou, Dr. Yang 邹杨
Yan, Dr. Rui
Xu, Prof. Hongjie
摘要： Molten salt reactors, being the only reactor type among Generation IV advanced nuclear reactors to utilize liquid fuel, offer inherent safety, high-temperature and low-pressure operation, as well as the capability for online fuel reprocessing. However, fuel salt flow results in the decay of delayed neutron precursors (DNPs) outside the core, leading to fluctuations in the effective delayed neutron fraction and consequently impacting reactor reactivity. Particularly under accident scenarios—such as combined pump shutdown and inability to rapidly scram the reactor—the reliance solely on negative temperature feedback may cause a substantial increase in core temperature, posing a threat to reactor safety. To address these issues, this paper introduces an innovative design for a passive fluid-driven Suspended Control Rod (SCR) aimed at dynamically compensating for reactivity fluctuations caused by DNPs flowing with fuel flow. The control rod operates passively by leveraging the combined effects of gravity, buoyancy, and fluid dynamic forces, thereby eliminating the need for any external drive mechanism and allowing direct integration within the core’s active region. Using a 150 MWth thorium-based molten salt reactor as the reference design, a mathematical model was developed to systematically analyze the effects of key parameters—including the SCR's geometric dimensions and density—on its performance, examine its motion characteristics under different core flow conditions, and assess its feasibility for dynamic compensation of reactivity changes caused by fuel flow. The study’s results demonstrate that the SCR can effectively counteract the reactivity fluctuations induced by fuel flow within molten salt reactors. Sensitivity analysis revealed that the SCR’s average density exerts a profound impact on its start-up flow threshold, channel flow rate, resistance to fuel density fluctuations, and response characteristics, underscoring the critical need to optimize this parameter. Moreover, by judiciously selecting the SCR’s length, number of deployed units, and placement, one can achieve the necessary reactivity control while also maintaining a favorable balance between neutron economy and heat transfer performance. Ultimately, this study provides an innovative solution for passive reactivity control in molten salt reactors, offering substantial potential for practical engineering applications.
分类： 物理学 >> 核物理学 提交时间： 2025-02-20
Yu, Mr. Changqing
Zhu, Dr. Guifeng
Jia, Dr. Shuyang
Zou, Dr. Yang 邹杨
Yan, Dr. Rui
Guo, Dr. Jian
Liu, Dr. Yafen
Zhou, Dr. Bo
Zhao, Dr. Xuechao
摘要： Knowing the precise relationship between fuel loading and reactivity helps guide the smooth progress of reactor criticality extrapolation and online refueling in molten salt reactors (MSRs). This study aims to explore and explain the linear relationship between reactivity and the reciprocal of uranium concentration in thermal spectrum MSRs. By applying the neutron balance theory, we analyzed the absorption of neutrons by various nuclides under several single lattice models with varying fuel concentrations. Our findings reveal a simple linear correlation between reactivity and the reciprocal of uranium concentration, which is successfully explained from the perspective of nuclear reaction cross-sections that adhere to the 1/v law in a thermal neutron spectrum. Furthermore, we identified the single-group neutron absorption cross-sections of structural materials and carrier salts exhibit an approximate linear relationship with the single-group fission cross-section of 235U, and the reciprocal of the fission cross-section of 235U exhibits an approximate linear relationship with uranium concentration. This linear relationship will deviate as the volume fraction of molten salt continues to increase since more neutrons will be captured in the resonance energy spectrum. But it remains valid within a 25% volume fraction of molten salt, and still demonstrates its broad applicability in the physical design and operation of thermal molten salt reactors.
分类： 物理学 >> 核物理学 提交时间： 2024-12-17
Yu, Mr. Changqing
Zhu, Dr. Guifeng
Jia, Dr. Shuyang
Zou, Dr. Yang 邹杨
Yan, Dr. Rui
Guo, Dr. Jian
Liu, Dr. Yafen
Zhou, Dr. Bo
Zhao, Dr. Xuechao
摘要： Knowing the precise relationship between fuel loading and reactivity helps guide the smooth progress of reactor criticality extrapolation and online refueling in molten salt reactors (MSRs). This study aims to explore and explain the linear relationship between reactivity and the reciprocal of uranium concentration in thermal spectrum MSRs. By applying the neutron balance theory, we analyzed neutron absorption by various nuclides under several single lattice models with varying fuel concentrations. Our findings reveal a simple linear correlation between reactivity and the reciprocal of uranium concentration, which is successfully explained from the perspective of nuclear reaction cross-sections that adhere to the 1/v law in a thermal neutron spectrum. Furthermore, we identified the single-group neutron absorption cross-sections of structural materials and carrier salts exhibit an approximate linear relationship with the single-group fission cross-section of 235U, and the reciprocal of the fission cross-section of 235U exhibits an approximate linear relationship with uranium concentration. This linear relationship will deviate as the volume fraction of molten salt continues to increase since more neutron will be captured in the resonance energy spectrum. But it remains valid within a 25% volume fraction of molten salt, and still demonstrates its broad applicability in the physical design and operation of thermal molten salt reactors.
分类： 物理学 >> 核物理学 提交时间： 2025-05-15
Li, Mr. Qi-Ming
Fu, Dr. Yao
Tian, Dr. Jian
Zhou, Dr. Chong
Zou, Dr. Yang 邹杨
fu, Prof. yuan 傅远
Hou, Dr. Jie 后接
Yu, Prof. Xiaohan
Xu, Prof. Hongjie
摘要： This study experimentally investigated the heat transfer performance of a novel shell-and-tube fluoride Salt to Salt Heat Exchanger (SSHX) featuring baffles with integrated drainage ports, designed to mitigate salt freeze blockage risks during shutdown in Molten Salt Reactors (MSRs). Experiments were conducted in a Scaled Simulation Fluoride Salt-cooled Reactor (SF0) test facility. A new empirical correlation for tube-side heat transfer was proposed as Nu=0.0246Re0.8Pr0.267 (valid for Re=9000~15000 and Pr=8~12), demonstrating excellent agreement with experimental data within a maximum deviation of 5%. Comparative analysis revealed the modified Dittus-Boelter equation is still a suitable choice for predicting fluoride salt convective heat transfer behavior in tubular geometries, outperforming the Gnielinski and Sieder-Tate models, which overpredicted data by 17-25%. For shell-side heat transfer, applying a 31% enhancement factor (ε=1.31) to the Kern correlation aligns predictions with experimental results within an error range of -6.0% to 7.0%. These findings address a critical engineering challenge in SSHXs while preserving thermal efficiency, offering essential experimental data and valuable insights for the design of fluoride SSHXs in MSRs.
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