Pub Date : 2024-11-20DOI: 10.1016/j.anucene.2024.111058
M.M.R. Williams
We demonstrate that the concept of a one-dimensional stochastic problem, in which only the statistical properties of the medium in one direction are used, is an unphysical situation. Even though the statistically averaged quantities, such as mean value and covariance, may depend only on one space dimension, the statistical properties of the medium in the other two directions must be included. A simple example, based on a second order differential equation, is used to illustrate the point and is supported by numerical calculations. The relevance of this matter to radiation and neutron transport in spatially stochastic media is made clear.
{"title":"The influence of spatially anisotropic randomness on the solution of one-dimensional stochastic differential and integral equations","authors":"M.M.R. Williams","doi":"10.1016/j.anucene.2024.111058","DOIUrl":"10.1016/j.anucene.2024.111058","url":null,"abstract":"<div><div>We demonstrate that the concept of a one-dimensional stochastic problem, in which only the statistical properties of the medium in one direction are used, is an unphysical situation. Even though the statistically averaged quantities, such as mean value and covariance, may depend only on one space dimension, the statistical properties of the medium in the other two directions must be included. A simple example, based on a second order differential equation, is used to illustrate the point and is supported by numerical calculations. The relevance of this matter to radiation and neutron transport in spatially stochastic media is made clear.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111058"},"PeriodicalIF":1.9,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704095","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-20DOI: 10.1016/j.anucene.2024.111056
Boubacar Kirgni Hamza , Wang Junling , Lasseini Gonga Yahaya Abdoul Razak , Moussa Hassane Ayouba
NPP is a complex time varying nonlinear system that have to operate under severe constraints while complying with safe operating conditions in order to ensure the power demand and prevent the plant from contingent network instability. Towards this goal, this paper proposes the design of active fault tolerant proportional and derivative (PD) control law for Pressurized Water Reactors (PWRs) under external disturbance. To achieve this, the reactor core’s nonlinear dynamic is transformed into input/output (I/O) second-order system with respect to the power level. Based on the new model, a control strategy is proposed that aims to detect, identify, estimate and compensate for actuator faults. The stability of the system is proven using Lyapunov stability theory, where a quadruple Linear Matrix Inequalities (LMIs) system is derived to provide gains for both the control law and observer. Numerical simulations are conducted to assess the performance of the newly built control strategy and a comparison has been made. It follows that the designed controller not only effectively manages faults and external disturbances but also demonstrate exceptional performance when compared to the model-free controller (MFC) and the L1 adaptive robust controller (L1 ARC).
{"title":"Active fault tolerant PD control law for PWRs under external disturbance","authors":"Boubacar Kirgni Hamza , Wang Junling , Lasseini Gonga Yahaya Abdoul Razak , Moussa Hassane Ayouba","doi":"10.1016/j.anucene.2024.111056","DOIUrl":"10.1016/j.anucene.2024.111056","url":null,"abstract":"<div><div>NPP is a complex time varying nonlinear system that have to operate under severe constraints while complying with safe operating conditions in order to ensure the power demand and prevent the plant from contingent network instability. Towards this goal, this paper proposes the design of active fault tolerant proportional and derivative (PD) control law for Pressurized Water Reactors (PWRs) under external disturbance. To achieve this, the reactor core’s nonlinear dynamic is transformed into input/output (I/O) second-order system with respect to the power level. Based on the new model, a control strategy is proposed that aims to detect, identify, estimate and compensate for actuator faults. The stability of the system is proven using Lyapunov stability theory, where a quadruple Linear Matrix Inequalities (LMIs) system is derived to provide gains for both the control law and observer. Numerical simulations are conducted to assess the performance of the newly built control strategy and a comparison has been made. It follows that the designed controller not only effectively manages faults and external disturbances but also demonstrate exceptional performance when compared to the model-free controller (MFC) and the L<sub>1</sub> adaptive robust controller (L<sub>1</sub> ARC).</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111056"},"PeriodicalIF":1.9,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704093","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
In Japan, critical assemblies are regulated under the “Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactions” which provides guidelines for the construction and the operation of nuclear facilities: the core configurations of nuclear experiments are constrained accordingly. In addition, the nuclear facilities are required to comply with strict administrations if the nuclear facilities have more than the certain amount of nuclear fuels. On the other hand, subcritical assemblies are not subjected to the nuclear reactor regulation because the subcritical assemblies are not classified as nuclear reactors. We propose the subcritical assembly with a pulsed accelerator neutron source, which uses nuclear fuels below the regulation amount to ease maintenance and to widen experimental capabilities. Conceptual study of such system was carried out to calculate the deviation between reactivity difference in critical state and that in subcritical state by using the MCNP 6.2 code and the JENDL 4 library. Neutron source multiplication method and pulsed neutron source method were performed. The deviation can be estimated around 10 % by setting detectors in appropriate positions. Additionally, neutron radiography was studied as a simple application. The geometrical shapes of sample rods were seen clearly.
{"title":"Conceptual study of subcritical assembly with neutron generator for reactor physics experiments and neutron utilization","authors":"Hiroshi Nakagomi, Kenichi Yoshioka, Tsukasa Sugita, Haruo Miyadera","doi":"10.1016/j.anucene.2024.111039","DOIUrl":"10.1016/j.anucene.2024.111039","url":null,"abstract":"<div><div>In Japan, critical assemblies are regulated under the “Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactions” which provides guidelines for the construction and the operation of nuclear facilities: the core configurations of nuclear experiments are constrained accordingly. In addition, the nuclear facilities are required to comply with strict administrations if the nuclear facilities have more than the certain amount of nuclear fuels. On the other hand, subcritical assemblies are not subjected to the nuclear reactor regulation because the subcritical assemblies are not classified as nuclear reactors. We propose the subcritical assembly with a pulsed accelerator neutron source, which uses nuclear fuels below the regulation amount to ease maintenance and to widen experimental capabilities. Conceptual study of such system was carried out to calculate the deviation between reactivity difference in critical state and that in subcritical state by using the MCNP 6.2 code and the JENDL 4 library. Neutron source multiplication method and pulsed neutron source method were performed. The deviation can be estimated around 10 % by setting detectors in appropriate positions. Additionally, neutron radiography was studied as a simple application. The geometrical shapes of sample rods were seen clearly.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111039"},"PeriodicalIF":1.9,"publicationDate":"2024-11-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704092","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-19DOI: 10.1016/j.anucene.2024.111004
Taikun Guo , Junying Hong , Rui Han , Ruifeng Tian , Sichao Tan , Jiming Wen
Direct contact condensation of steam bubbles with non-condensable gas is common in nuclear safety equipment. It was insufficient research on the direct contact condensation heat transfer model of steam bubbles with non-condensable gas at high Reynolds numbers (Re > 22000) in previous works. To study the effect of non-condensable gas on heat transfer of steam–air bubbles at high Reynolds numbers, a high-speed camera was used to capture the behavior of bubbles and used image processing and bubble reconstruction to obtain size and dynamic parameters of bubbles. The size and upward motion behavior of bubbles were analyzed. The heat transfer coefficient during the bubbles condensation experiment were calculated and experimental results were compared with correlations proposed by previous works. In order to better explain the heat transfer characteristics and predict the heat transfer coefficient of bubbles at high Reynolds numbers, a modified heat transfer correlation based on the correlation of rigid sphere was proposed which are functions of bubble Reynolds number, liquid Prandtl number, Jacob number, and dimensionless time. This correlation considers the influence of both forced convection around bubbles and variations in steam fractions on bubble condensation. Comparison of experimental data and the corresponding predicted values shows that the deviation between the experimental data and predicted values is within ± 25 % which indicates the modified correlation accurately predicts the experimental data in this paper.
{"title":"Experimental study on direct contact condensation of steam bubbles with non-condensable gas at high Reynolds numbers","authors":"Taikun Guo , Junying Hong , Rui Han , Ruifeng Tian , Sichao Tan , Jiming Wen","doi":"10.1016/j.anucene.2024.111004","DOIUrl":"10.1016/j.anucene.2024.111004","url":null,"abstract":"<div><div>Direct contact condensation of steam bubbles with non-condensable gas is common in nuclear safety equipment. It was insufficient research on the direct contact condensation heat transfer model of steam bubbles with non-condensable gas at high Reynolds numbers (<em>Re</em> > 22000) in previous works. To study the effect of non-condensable gas on heat transfer of steam–air bubbles at high Reynolds numbers, a high-speed camera was used to capture the behavior of bubbles and used image processing and bubble reconstruction to obtain size and dynamic parameters of bubbles. The size and upward motion behavior of bubbles were analyzed. The heat transfer coefficient during the bubbles condensation experiment were calculated and experimental results were compared with correlations proposed by previous works. In order to better explain the heat transfer characteristics and predict the heat transfer coefficient of bubbles at high Reynolds numbers, a modified heat transfer correlation based on the correlation of rigid sphere was proposed which are functions of bubble Reynolds number, liquid Prandtl number, Jacob number, and dimensionless time. This correlation considers the influence of both forced convection around bubbles and variations in steam fractions on bubble condensation. Comparison of experimental data and the corresponding predicted values shows that the deviation between the experimental data and predicted values is within ± 25 % which indicates the modified correlation accurately predicts the experimental data in this paper.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111004"},"PeriodicalIF":1.9,"publicationDate":"2024-11-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142704091","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-16DOI: 10.1016/j.anucene.2024.111052
Xiaobo Li , Xunchao Zhang , Yuanshuai Qin , Yuan He
Integrated target module with a solid beam window, and cooled by reactor primary coolant is a good contender for Accelerator Driven System (ADS) and the cooling of the beam window is a key technique in it. The numerical analysis of two beam profiles (circular/double-circular scan) for the target assembly was performed by computational fluid dynamics (CFD) method, and a combing method was used to optimize the thermal–hydraulic design. The calculation results indicated that a nozzle was required to lower the maximum external surface temperature of the beam window to below 400℃. This can be achieved by reducing the heat deposited densities or increasing the velocity in the main heat deposited zone without enhancing the heat convection at the window surface, this will effectively reduce the temperature, but without increasing the temperature difference in the beam window. The optimization results indicate that the maximum temperature of the beam window is 393.75℃ and 384.94℃ for the circular scan and the double-circular scan, respectively. Additionally, the maximum temperature difference is 23.0℃ and 17.1℃ for the circular scan and the double-circular scan, respectively.
{"title":"Multi-Beam Accelerator-Driven-Systems of Part I: Optimization on Thermal-Hydraulic design of target assembly","authors":"Xiaobo Li , Xunchao Zhang , Yuanshuai Qin , Yuan He","doi":"10.1016/j.anucene.2024.111052","DOIUrl":"10.1016/j.anucene.2024.111052","url":null,"abstract":"<div><div>Integrated target module with a solid beam window, and cooled by reactor primary coolant is a good contender for Accelerator Driven System (ADS) and the cooling of the beam window is a key technique in it. The numerical analysis of two beam profiles (circular/double-circular scan) for the target assembly was performed by computational fluid dynamics (CFD) method, and a combing method was used to optimize the thermal–hydraulic design. The calculation results indicated that a nozzle was required to lower the maximum external surface temperature of the beam window to below 400℃. This can be achieved by reducing the heat deposited densities or increasing the velocity in the main heat deposited zone without enhancing the heat convection at the window surface, this will effectively reduce the temperature, but without increasing the temperature difference in the beam window. The optimization results indicate that the maximum temperature of the beam window is 393.75℃ and 384.94℃ for the circular scan and the double-circular scan, respectively. Additionally, the maximum temperature difference is 23.0℃ and 17.1℃ for the circular scan and the double-circular scan, respectively.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111052"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658314","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-16DOI: 10.1016/j.anucene.2024.111059
M. Garcia, L.E. Herranz
This study focuses on the release of the most relevant volatile fission products from the hot sodium pools. From an exhaustive review of the open literature, a critical review of a model based on the film theory and its estimates in comparison to data from the NALA experimental program has been conducted. Due to certain assumptions embedded in the approach, the fitting-to-data transport coefficients derived and some inconsistencies found between its formulation and the reported estimates, an alternate approach including other phenomena anticipated in the scenario is here proposed. Based on diffusive and convective mechanisms in the gas phase, the assumption of the analogy between heat and mass transport (HMT) and the ideal-dilute solutions laws to set the fission products concentration at the Na pool interface, a good agreement has been found with experimental data, which mean a substantial enhancement of qualitative and quantitative predictability while maintaining a conservative nature.
本研究的重点是热钠池中最相关的挥发性裂变产物的释放。通过详尽查阅公开文献，对基于薄膜理论的模型及其与 NALA 实验项目数据的估计值进行了严格审查。由于该方法中的某些假设、所得出的与数据相匹配的传输系数以及其表述与所报告的估计值之间的一些不一致之处，我们在此提出了一种替代方法，其中包括方案中预期的其他现象。基于气相中的扩散和对流机制、热量和质量输运（HMT）之间的类比假设以及理想稀释解法来设定 Na 池界面的裂变产物浓度，我们发现该方法与实验数据非常吻合，这意味着在保持保守性的同时，定性和定量预测能力得到了大幅提升。
{"title":"Volatile fission products transfer from hot sodium pools","authors":"M. Garcia, L.E. Herranz","doi":"10.1016/j.anucene.2024.111059","DOIUrl":"10.1016/j.anucene.2024.111059","url":null,"abstract":"<div><div>This study focuses on the release of the most relevant volatile fission products from the hot sodium pools. From an exhaustive review of the open literature, a critical review of a model based on the film theory and its estimates in comparison to data from the NALA experimental program has been conducted. Due to certain assumptions embedded in the approach, the fitting-to-data transport coefficients derived and some inconsistencies found between its formulation and the reported estimates, an alternate approach including other phenomena anticipated in the scenario is here proposed. Based on diffusive and convective mechanisms in the gas phase, the assumption of the analogy between heat and mass transport (HMT) and the ideal-dilute solutions laws to set the fission products concentration at the Na pool interface, a good agreement has been found with experimental data, which mean a substantial enhancement of qualitative and quantitative predictability while maintaining a conservative nature.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111059"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658347","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-16DOI: 10.1016/j.anucene.2024.111057
Shusong Qin , Binxian He , Xiangfei Meng , Jianchuang Sun , Wenchao Zhang , Weihua Cai
As a new type of fuel assembly, the thermo-hydraulic and mechanical properties of petal-shaped fuel rod are directly related to the safe operation of the reactor. In this study, the multi-physics field coupled model for coolant single-phase flow heat transfer, and mechanical properties of three-petal fuel rod is established through ABAQUS-STAR CCM+, realizing real-time data interaction between different computational domains. The results show that the transverse flow of coolant affects the temperature field distribution, and the multi-physics field coupling is closer to the real situation of convective heat transfer of fuel rods. Under the irradiation swelling, the maximum Mises stress moved from the inner concave arc to the outer convex arc, and increased to 433.23 MPa at the burnup of 8.22 % fissions of initial mental atoms (FIMA). In addition, the properties of the fuel rod at different inlet flow velocities are analyzed. When the burnup reaches 3.31 % FIMA at 2.5 m/s inlet flow velocity, it enters the plastic stage earlier. The effect of different coupling modes on displacement deformation is also discussed, which shows that it is feasible to ignore the influence of displacement deformation of the fuel rod on coolant.
{"title":"Study on single-phase flow and irradiation mechanics multi-physics field coupling properties of three-petal fuel rod","authors":"Shusong Qin , Binxian He , Xiangfei Meng , Jianchuang Sun , Wenchao Zhang , Weihua Cai","doi":"10.1016/j.anucene.2024.111057","DOIUrl":"10.1016/j.anucene.2024.111057","url":null,"abstract":"<div><div>As a new type of fuel assembly, the thermo-hydraulic and mechanical properties of petal-shaped fuel rod are directly related to the safe operation of the reactor. In this study, the multi-physics field coupled model for coolant single-phase flow heat transfer, and mechanical properties of three-petal fuel rod is established through ABAQUS-STAR CCM+, realizing real-time data interaction between different computational domains. The results show that the transverse flow of coolant affects the temperature field distribution, and the multi-physics field coupling is closer to the real situation of convective heat transfer of fuel rods. Under the irradiation swelling, the maximum Mises stress moved from the inner concave arc to the outer convex arc, and increased to 433.23 MPa at the burnup of 8.22 % fissions of initial mental atoms (FIMA). In addition, the properties of the fuel rod at different inlet flow velocities are analyzed. When the burnup reaches 3.31 % FIMA at 2.5 m/s inlet flow velocity, it enters the plastic stage earlier. The effect of different coupling modes on displacement deformation is also discussed, which shows that it is feasible to ignore the influence of displacement deformation of the fuel rod on coolant.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111057"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658312","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-16DOI: 10.1016/j.anucene.2024.111021
Evan S. Gonzalez , Brian C. Kiedrowski , Gregory G. Davidson
The Transient Multi-Level (TML) method is applied to a time-dependent Monte Carlo transport solver to offload some of the computational burden of the expensive Monte Carlo solve to lower-order Coarse Mesh Finite Difference (CMFD) and Exact Point Kinetics Equations (EPKE) solvers via factorization of the neutron flux at the transport and CMFD levels using the Predictor Corrector Quasi-Static Method (PCQM). The Monte Carlo transient is solved by a modified fission source iteration scheme that introduces a single transient source bank. The method is implemented in the production-level Monte Carlo code, Shift, and verified with prescribed reactivity ramps from the two-dimensional version of the C5G7-TD reactor benchmark. The results show that, as compared to other quasi-static methods, the TML reduces the stochastic noise inherent to the transient Monte Carlo solver by factors of 2 to 6 for various norm comparisons of the reactor power amplitude. The TML additionally reduces the number of Monte Carlo evaluations needed to simulate the transient, leading to roughly an order of magnitude improvement in CPU time relative to the standard PCQM for the problems tested.
{"title":"The Transient Multi-Level method for Monte Carlo reactor statics calculations","authors":"Evan S. Gonzalez , Brian C. Kiedrowski , Gregory G. Davidson","doi":"10.1016/j.anucene.2024.111021","DOIUrl":"10.1016/j.anucene.2024.111021","url":null,"abstract":"<div><div>The Transient Multi-Level (TML) method is applied to a time-dependent Monte Carlo transport solver to offload some of the computational burden of the expensive Monte Carlo solve to lower-order Coarse Mesh Finite Difference (CMFD) and Exact Point Kinetics Equations (EPKE) solvers via factorization of the neutron flux at the transport and CMFD levels using the Predictor Corrector Quasi-Static Method (PCQM). The Monte Carlo transient is solved by a modified fission source iteration scheme that introduces a single transient source bank. The method is implemented in the production-level Monte Carlo code, Shift, and verified with prescribed reactivity ramps from the two-dimensional version of the C5G7-TD reactor benchmark. The results show that, as compared to other quasi-static methods, the TML reduces the stochastic noise inherent to the transient Monte Carlo solver by factors of <span><math><mo>∼</mo></math></span>2 to 6 for various norm comparisons of the reactor power amplitude. The TML additionally reduces the number of Monte Carlo evaluations needed to simulate the transient, leading to roughly an order of magnitude improvement in CPU time relative to the standard PCQM for the problems tested.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111021"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658309","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-16DOI: 10.1016/j.anucene.2024.111050
Li Liu , Yamin Li , Hanyang Gu
Swirl-vane separator is essential in marine nuclear SG for separating liquid droplets from gas flow. This study conducts 3D numerical simulations to investigate the impact of rolling amplitude (θm) and rolling period (T) on flow field and separation performance under rolling motion, incorporating additional inertia forces through a UDF. Results indicate that non-axisymmetric patterns in vapor velocity and total pressure fields exhibit periodic fluctuations, lagging approximately one-quarter period behind the separator’s movement due to inertia forces. The fluctuation of liquid volume fraction resembles a sinusoidal periodic pattern. Separation efficiency fluctuates in an “M” pattern within 10° < θm < 40° and 4 s < T < 8 s, and in a “W” at θm = 40° and 30°, with T = 2 s. Pressure loss fluctuation follows a “W” pattern. Increasing rolling amplitude and decreasing period intensify liquid film thickness variability near separator walls, reducing separation efficiency, increasing pressure loss.
漩涡叶片分离器是船用核安全气囊中必不可少的分离气流中液滴的装置。本研究进行了三维数值模拟，研究滚动幅度（θm）和滚动周期（T）对滚动运动下流场和分离性能的影响，并通过 UDF 加入了额外的惯性力。结果表明，由于惯性力的作用，蒸汽速度和总压力场的非轴对称模式表现出周期性波动，大约滞后于分离器运动的四分之一周期。液体体积分数的波动类似于正弦周期模式。分离效率在 10° < θm < 40° 和 4 s < T < 8 s 内呈 "M "型波动，在 θm = 40° 和 30° 时呈 "W "型波动，T = 2 s。滚动幅度的增加和周期的减小加剧了分离器壁附近液膜厚度的变化，降低了分离效率，增加了压力损失。
{"title":"Numerical study on flow field and separation performance of swirl-vane separator under rolling motion conditions","authors":"Li Liu , Yamin Li , Hanyang Gu","doi":"10.1016/j.anucene.2024.111050","DOIUrl":"10.1016/j.anucene.2024.111050","url":null,"abstract":"<div><div>Swirl-vane separator is essential in marine nuclear SG for separating liquid droplets from gas flow. This study conducts 3D numerical simulations to investigate the impact of rolling amplitude (<em>θ</em><sub>m</sub>) and rolling period (<em>T</em>) on flow field and separation performance under rolling motion, incorporating additional inertia forces through a UDF. Results indicate that non-axisymmetric patterns in vapor velocity and total pressure fields exhibit periodic fluctuations, lagging approximately one-quarter period behind the separator’s movement due to inertia forces. The fluctuation of liquid volume fraction resembles a sinusoidal periodic pattern. Separation efficiency fluctuates in an “M” pattern within 10° < <em>θ</em><sub>m</sub> < 40° and 4 s < <em>T</em> < 8 s, and in a “W” at <em>θ</em><sub>m</sub> = 40° and 30°, with <em>T</em> = 2 s. Pressure loss fluctuation follows a “W” pattern. Increasing rolling amplitude and decreasing period intensify liquid film thickness variability near separator walls, reducing separation efficiency, increasing pressure loss.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111050"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658308","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-16DOI: 10.1016/j.anucene.2024.111049
Yue Lin , Dalin Zhang , Yutong Chen , Xisi Zhang , Jian Deng , Wenxi Tian , Suizheng Qiu , Guanghui Su
Melting and freezing model plays a crucial role in safety analysis for liquid metal fast reactors under core disruptive accidents. It’s necessary to establish suitable models for accurate prediction of molten materials migration after severe accidents happens. In this study, a kind of melting and freezing model is established in ACENA code. And the ACENA code is used to simulate the freezing behavior of melt and calculated results are compared with experimental results to validate the applicability of the model. According to the comparison it is shown that the simulation results by ACENA code agree well with experimental results. Both the changing trend and final value of penetration depth of melt are consistent with realistic conditions. Besides, the effects of heat exchange on penetration depth are also analyzed. This research contributes to the development of safety analysis tools for liquid metal reactors under severe accidents.
{"title":"Preliminary development and validation of ACENA code for melting and freezing behavior simulation","authors":"Yue Lin , Dalin Zhang , Yutong Chen , Xisi Zhang , Jian Deng , Wenxi Tian , Suizheng Qiu , Guanghui Su","doi":"10.1016/j.anucene.2024.111049","DOIUrl":"10.1016/j.anucene.2024.111049","url":null,"abstract":"<div><div>Melting and freezing model plays a crucial role in safety analysis for liquid metal fast reactors under core disruptive accidents. It’s necessary to establish suitable models for accurate prediction of molten materials migration after severe accidents happens. In this study, a kind of melting and freezing model is established in ACENA code. And the ACENA code is used to simulate the freezing behavior of melt and calculated results are compared with experimental results to validate the applicability of the model. According to the comparison it is shown that the simulation results by ACENA code agree well with experimental results. Both the changing trend and final value of penetration depth of melt are consistent with realistic conditions. Besides, the effects of heat exchange on penetration depth are also analyzed. This research contributes to the development of safety analysis tools for liquid metal reactors under severe accidents.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"212 ","pages":"Article 111049"},"PeriodicalIF":1.9,"publicationDate":"2024-11-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142658311","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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