Correction to:Nuclear Science and Techniques(2025)36:111 https://doi.org/10.1007/s41365-025-01681-9.In the sentence beginning‘The weights of the parameters used for the…’in this article,the text‘RCSs’should have ...Correction to:Nuclear Science and Techniques(2025)36:111 https://doi.org/10.1007/s41365-025-01681-9.In the sentence beginning‘The weights of the parameters used for the…’in this article,the text‘RCSs’should have read‘SCRs’.In Table 7 of this article,the column header ρ_fuel was incorrect and should have read CPv_fuel.For completeness and transparency,the old incorrect version and the corrected version of Table 7 are displayed below.展开更多
Knowing the precise relationship between fuel loading and reactivity is essential for guiding reactor criticality extrapolation and online refueling in molten salt reactors(MSRs).This study aims to explore and explain...Knowing the precise relationship between fuel loading and reactivity is essential for guiding 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 neutron balance theory,we analyzed the neutron absorption cross sections of various nuclides in single-lattice models with varying fuel concentrations.Our findings reveal a simple linear correlation between reactivity and the reciprocal of uranium concentration,which can be explained from the perspective of nuclear reaction cross sections that adhere to the 1/v law in the thermal neutron spectrum.Furthermore,we identified that the neutron absorption single-group cross sections of structural materials and carrier salts exhibit an approximately linear relationship with the fission single-group cross section of ^(235) U;similarly,the reciprocal of ^(235)U’s fission cross section exhibits an approximately linear relationship with uranium concentration.This linear relationship deviates as the volume fraction of molten salt increases,due to a greater proportion of neutrons being captured in the resonance energy spectrum.However,it remains valid for molten salt volume fractions up to 25%and demonstrates broad applicability in the physical design and operation of thermal molten salt reactors.展开更多
DAYU3D is a modern three-dimensional(3D)computer code for thermal-hydraulic design and accident analysis in hightemperature gas-cooled reactors(HTGRs),developed by the Institute of Nuclear and New Energy Technology(IN...DAYU3D is a modern three-dimensional(3D)computer code for thermal-hydraulic design and accident analysis in hightemperature gas-cooled reactors(HTGRs),developed by the Institute of Nuclear and New Energy Technology(INET)at Tsinghua University.Compared to the traditional codes like TINTE,the DAYU3D code has advantages due to its refined framework,improved models,and more efficient algorithms.It is able to simulate the continuous movement of control rods and is more rigorous in treating radiation heat transfer and the break mass flow.Advanced computational methods significantly improve the computational efficiency of DAYU3D,achieving a time reduction of over 60%compared to TINTE.Extensive verification and validation with more than 100 cases demonstrate that DAYU3D is promising for HTGR 3D thermal-hydraulic design and accident analyses.展开更多
Molten salt reactors,being the only reactor type among Generation Ⅳ advanced nuclear reactors that utilize liquid fuels,offer inherent safety,high-temperature,and low-pressure operation,as well as the capability for ...Molten salt reactors,being the only reactor type among Generation Ⅳ advanced nuclear reactors that utilize liquid fuels,offer inherent safety,high-temperature,and low-pressure operation,as well as the capability for online fuel reprocessing.However,the fuel-salt flow results in the decay of delayed neutron precursors(DNPs)outside the core,causing fluctuations in the effective delayed neutron fraction and consequently impacting the reactor reactivity.Particularly in accident scenarios—such as a combined pump shutdown and the inability to rapidly scram the reactor—the sole reliance on negative temperature feedback may cause a significant increase in core temperature,posing a threat to reactor safety.To address these problems,this paper introduces an innovative design for a passive fluid-driven suspended control rod(SCR)to dynamically compensate for reactivity fluctuations caused by DNPs flowing with the fuel.The control rod operates passively by leveraging the combined effects of gravity,buoyancy,and fluid dynamic forces,thereby eliminating the need for an external drive mechanism and enabling direct integration within the active region of the core.Using a 150 MWt thorium-based molten salt reactor as the reference design,we develop a mathematical model to systematically analyze the effects of key parameters—including the geometric dimensions and density of the SCR—on its performance.We examine its motion characteristics under different core flow conditions and assess its feasibility for the dynamic compensation of reactivity changes caused by fuel flow.The results of this study demonstrate that the SCR can effectively counteract reactivity fluctuations induced by fuel flow within molten salt reactors.A sensitivity analysis reveals 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.This underscores the critical need to optimize this parameter.Moreover,by judiciously selecting the SCR’s length,number of deployed units,and the placement we can achieve the necessary reactivity control while maintaining a favorable balance between neutron economy and heat transfer performance.Ultimately,this paper provides an innovative solution for the passive reactivity control in molten salt reactors,offering significant potential for practical engineering applications.展开更多
钠冷快堆(Sodium-cooled Fast Reactor,SFR)作为第四代核能系统的代表性堆型,凭借其燃料增殖潜力、次锕系核素嬗变能力及固有安全特性成为国际研究热点。然而,钠冷快堆设计与运行过程中普遍存在核数据随机性、模型简化误差、边界条件扰...钠冷快堆(Sodium-cooled Fast Reactor,SFR)作为第四代核能系统的代表性堆型,凭借其燃料增殖潜力、次锕系核素嬗变能力及固有安全特性成为国际研究热点。然而,钠冷快堆设计与运行过程中普遍存在核数据随机性、模型简化误差、边界条件扰动等多源不确定性,上述不确定性与反应堆内复杂的中子学行为、钠热工水力特性及事故工况下的多物理场效应耦合传播,可能导致钠冷快堆系统响应偏离或异常波动,甚至引发事故风险。本文主要从核反应堆物理计算、热工水力分析和事故安全分析三个方向概述了钠冷快堆系统不确定性分析的研究进展,总结了国际范围内的研究成果,分析了遇到的技术挑战和未来研究的发展趋势,并探讨了我国在这些领域的研究现状及未来的研究方向。通过分析与总结,希望为未来钠冷快堆系统的安全性评估与设计优化提供一定参考。展开更多
文摘Correction to:Nuclear Science and Techniques(2025)36:111 https://doi.org/10.1007/s41365-025-01681-9.In the sentence beginning‘The weights of the parameters used for the…’in this article,the text‘RCSs’should have read‘SCRs’.In Table 7 of this article,the column header ρ_fuel was incorrect and should have read CPv_fuel.For completeness and transparency,the old incorrect version and the corrected version of Table 7 are displayed below.
基金supported by the Youth Innovation Promotion Association of the Chinese Academy of Sciences(No.2020261)the Strategic Priority Research Program of the Chinese Academy of Sciences(No.XDA02010000)the Young Potential Program of the Shanghai Institute of Applied Physics,Chinese Academy of Sciences(No.SINAP-YXJH-202412)。
文摘Knowing the precise relationship between fuel loading and reactivity is essential for guiding 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 neutron balance theory,we analyzed the neutron absorption cross sections of various nuclides in single-lattice models with varying fuel concentrations.Our findings reveal a simple linear correlation between reactivity and the reciprocal of uranium concentration,which can be explained from the perspective of nuclear reaction cross sections that adhere to the 1/v law in the thermal neutron spectrum.Furthermore,we identified that the neutron absorption single-group cross sections of structural materials and carrier salts exhibit an approximately linear relationship with the fission single-group cross section of ^(235) U;similarly,the reciprocal of ^(235)U’s fission cross section exhibits an approximately linear relationship with uranium concentration.This linear relationship deviates as the volume fraction of molten salt increases,due to a greater proportion of neutrons being captured in the resonance energy spectrum.However,it remains valid for molten salt volume fractions up to 25%and demonstrates broad applicability in the physical design and operation of thermal molten salt reactors.
文摘DAYU3D is a modern three-dimensional(3D)computer code for thermal-hydraulic design and accident analysis in hightemperature gas-cooled reactors(HTGRs),developed by the Institute of Nuclear and New Energy Technology(INET)at Tsinghua University.Compared to the traditional codes like TINTE,the DAYU3D code has advantages due to its refined framework,improved models,and more efficient algorithms.It is able to simulate the continuous movement of control rods and is more rigorous in treating radiation heat transfer and the break mass flow.Advanced computational methods significantly improve the computational efficiency of DAYU3D,achieving a time reduction of over 60%compared to TINTE.Extensive verification and validation with more than 100 cases demonstrate that DAYU3D is promising for HTGR 3D thermal-hydraulic design and accident analyses.
基金supported by Youth Innovation Promotion Association of Chinese Academy of Sciences(No.2020261)Strategic Priority Research Program of Chinese Academy of Sciences(No.XDA02010000)the Young Potential Program of Shanghai Institute of Applied Physics,Chinese Academy of Sciences(No.SINAP-YXJH-202412).
文摘Molten salt reactors,being the only reactor type among Generation Ⅳ advanced nuclear reactors that utilize liquid fuels,offer inherent safety,high-temperature,and low-pressure operation,as well as the capability for online fuel reprocessing.However,the fuel-salt flow results in the decay of delayed neutron precursors(DNPs)outside the core,causing fluctuations in the effective delayed neutron fraction and consequently impacting the reactor reactivity.Particularly in accident scenarios—such as a combined pump shutdown and the inability to rapidly scram the reactor—the sole reliance on negative temperature feedback may cause a significant increase in core temperature,posing a threat to reactor safety.To address these problems,this paper introduces an innovative design for a passive fluid-driven suspended control rod(SCR)to dynamically compensate for reactivity fluctuations caused by DNPs flowing with the fuel.The control rod operates passively by leveraging the combined effects of gravity,buoyancy,and fluid dynamic forces,thereby eliminating the need for an external drive mechanism and enabling direct integration within the active region of the core.Using a 150 MWt thorium-based molten salt reactor as the reference design,we develop a mathematical model to systematically analyze the effects of key parameters—including the geometric dimensions and density of the SCR—on its performance.We examine its motion characteristics under different core flow conditions and assess its feasibility for the dynamic compensation of reactivity changes caused by fuel flow.The results of this study demonstrate that the SCR can effectively counteract reactivity fluctuations induced by fuel flow within molten salt reactors.A sensitivity analysis reveals 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.This underscores the critical need to optimize this parameter.Moreover,by judiciously selecting the SCR’s length,number of deployed units,and the placement we can achieve the necessary reactivity control while maintaining a favorable balance between neutron economy and heat transfer performance.Ultimately,this paper provides an innovative solution for the passive reactivity control in molten salt reactors,offering significant potential for practical engineering applications.
文摘钠冷快堆(Sodium-cooled Fast Reactor,SFR)作为第四代核能系统的代表性堆型,凭借其燃料增殖潜力、次锕系核素嬗变能力及固有安全特性成为国际研究热点。然而,钠冷快堆设计与运行过程中普遍存在核数据随机性、模型简化误差、边界条件扰动等多源不确定性,上述不确定性与反应堆内复杂的中子学行为、钠热工水力特性及事故工况下的多物理场效应耦合传播,可能导致钠冷快堆系统响应偏离或异常波动,甚至引发事故风险。本文主要从核反应堆物理计算、热工水力分析和事故安全分析三个方向概述了钠冷快堆系统不确定性分析的研究进展,总结了国际范围内的研究成果,分析了遇到的技术挑战和未来研究的发展趋势,并探讨了我国在这些领域的研究现状及未来的研究方向。通过分析与总结,希望为未来钠冷快堆系统的安全性评估与设计优化提供一定参考。
