In this study, a thermal-hydraulic analysis is carried out for the core of natural circulation lead-cooled fast reactor SNCLFR-100. Steady-state and transient analysis are performed with porous medium approach-based code TWOPORFLOW. In the steady-state analysis, mass flow distribution and temperature distributions of the assemblies are analyzed in assembly-wise mode. At the same time, the hottest assembly is analyzed in pin level and the safety performance is investigated. In the transient analysis, a typical Design Extension Conditions unprotected over-power transient is simulated with one-way coupling method.
{"title":"Thermal-Hydraulic Safety Analysis of Natural Circulation Lead-Cooled Fast Reactor SNCLFR-100 Core Based on Porous Medium Approach","authors":"Wenpei Feng, Guangliang Yang, Kefan Zhang, Chong Qin, Xiao-mei Luo, Tao Ding, Hongli Chen","doi":"10.1115/icone29-92379","DOIUrl":"https://doi.org/10.1115/icone29-92379","url":null,"abstract":"\u0000 In this study, a thermal-hydraulic analysis is carried out for the core of natural circulation lead-cooled fast reactor SNCLFR-100. Steady-state and transient analysis are performed with porous medium approach-based code TWOPORFLOW. In the steady-state analysis, mass flow distribution and temperature distributions of the assemblies are analyzed in assembly-wise mode. At the same time, the hottest assembly is analyzed in pin level and the safety performance is investigated. In the transient analysis, a typical Design Extension Conditions unprotected over-power transient is simulated with one-way coupling method.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"78 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115314947","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
To adapt to a requirement of improving the accuracy and efficiency of calculation, a full or partial implicit scheme is usually employed in solving the conservative equations of the supercritical carbon dioxide (S-CO2) Brayton cycle, and partial derivatives of thermal properties such as (∂h/∂ρ)p and (∂h/∂p)ρ are needed in numerical solver. In this paper, the most representative state equations of carbon dioxide are investigated and evaluated by experimental data. The Span-Wagner (SW) equation has a minimal error in all state equations, so the SW equation is chosen as the fundamental equation of thermal properties for partial derivatives. Based on that, the equations of partial derivatives such as (∂h/∂ρ)p and (∂h/∂p)ρ are presented by the Maxwell equation. The paper also evaluates the closure of partial derivatives equations. The deviations of (∂h/∂ρ)p and (∂h/∂p)ρ are within ±0.01% for most points. The maximum closure error of (∂h/∂ρ)p is 0.373%, and the maximum one of (∂h/∂p)ρ is −0.798%. Therefore, the partial derivatives equations obtained in this paper can play a significant role in the safety analysis code.
{"title":"Assessment on Partial Derivatives for Thermal-Physical Properties of Carbon Dioxide","authors":"Shuang Wen, Q. Wen","doi":"10.1115/icone29-93314","DOIUrl":"https://doi.org/10.1115/icone29-93314","url":null,"abstract":"\u0000 To adapt to a requirement of improving the accuracy and efficiency of calculation, a full or partial implicit scheme is usually employed in solving the conservative equations of the supercritical carbon dioxide (S-CO2) Brayton cycle, and partial derivatives of thermal properties such as (∂h/∂ρ)p and (∂h/∂p)ρ are needed in numerical solver. In this paper, the most representative state equations of carbon dioxide are investigated and evaluated by experimental data. The Span-Wagner (SW) equation has a minimal error in all state equations, so the SW equation is chosen as the fundamental equation of thermal properties for partial derivatives. Based on that, the equations of partial derivatives such as (∂h/∂ρ)p and (∂h/∂p)ρ are presented by the Maxwell equation. The paper also evaluates the closure of partial derivatives equations. The deviations of (∂h/∂ρ)p and (∂h/∂p)ρ are within ±0.01% for most points. The maximum closure error of (∂h/∂ρ)p is 0.373%, and the maximum one of (∂h/∂p)ρ is −0.798%. Therefore, the partial derivatives equations obtained in this paper can play a significant role in the safety analysis code.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116535355","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
With the growing requirement of predicting reactor behavior in high-fidelity detail at practical conditions, it is urgent to accomplish thermal hydraulics (T-H) feedback in the high-fidelity neutron transport program HNET. For better convergence behaviors than Picard iteration, the Matrix Free Newton/Krylov (MFNK) method was employed to resolve neutronics and thermal-hydraulics coupling system. MFNK treats each subsystem as a black box within the Newton method framework, so it can facilitate the coupling procedure without surrendering efficiency or robustness. For the T-H feedback effects, a simplified internal thermal hydraulics model was adopted to provide T-H conditions for neutronics. The convergence behaviors of MFNK and Picard iteration were assessed through simple typical cases. Finally, the effectiveness of the coupling system was verified by the VERA problem #6. Numerical results demonstrate the efficiency and stability of MFNK compared with Picard iteration. Moreover, it turns out that the coupling system has a good performance in realistic cases.
{"title":"Implementation and Analysis of Thermal Hydraulics Feedback in High-Fidelity Neutron Transport Program HNET","authors":"Yanling Zhu, Chen Hao, Peijun Li, Xiaoyu Zhou","doi":"10.1115/icone29-92724","DOIUrl":"https://doi.org/10.1115/icone29-92724","url":null,"abstract":"\u0000 With the growing requirement of predicting reactor behavior in high-fidelity detail at practical conditions, it is urgent to accomplish thermal hydraulics (T-H) feedback in the high-fidelity neutron transport program HNET. For better convergence behaviors than Picard iteration, the Matrix Free Newton/Krylov (MFNK) method was employed to resolve neutronics and thermal-hydraulics coupling system. MFNK treats each subsystem as a black box within the Newton method framework, so it can facilitate the coupling procedure without surrendering efficiency or robustness. For the T-H feedback effects, a simplified internal thermal hydraulics model was adopted to provide T-H conditions for neutronics. The convergence behaviors of MFNK and Picard iteration were assessed through simple typical cases. Finally, the effectiveness of the coupling system was verified by the VERA problem #6. Numerical results demonstrate the efficiency and stability of MFNK compared with Picard iteration. Moreover, it turns out that the coupling system has a good performance in realistic cases.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"136 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116721562","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Two-phase flow measurement with high precision plays a significant role in safe and efficient operation of nuclear reactor. This paper focuses on the void fraction of gas-liquid flow in a non-transparent tube with the inner diameter of 13mm. X-ray computer tomography (CT) is considered to be effective for two-phase flow measurement because of its good qualities of non-invasion. However, due to the mechanical limitation of rotating speed, solutions for fast CT system to reduce the scanning time have been suggested relied on multiple sources and detectors. Recently, carbon nanotube (CNT) based X-ray source has been applied to CT imaging, significantly improves the temporal resolution by increasing the number of sources and avoid the gantry rotation. This paper proposes a potential static CT system design for the imaging of two-phase flow in straight steel tube. The setup of this system employed 90 couples of CNT X-ray sources and detector arrays arranged in a circle. Gas-liquid flow was simulated with different sizes of spheroidic bubbles randomly placed in the water inside the tube. To substitute for the flow moving, the z-axis of the simulator was added according to the flow velocity and exposure duration. Iterative image reconstruction was applied for the inverse problem of density distribution, and the reconstruction result of the experiment indicates the static CT system is useful to distinguish the flow pattern and measure the void faction distribution of gas-liquid flow.
{"title":"Measurement of Two-Phase Flow: Static CT System Based on Carbon Nanotubes","authors":"Yucheng Zhang, Shuo Xu, Xincheng Xiang","doi":"10.1115/icone29-92492","DOIUrl":"https://doi.org/10.1115/icone29-92492","url":null,"abstract":"\u0000 Two-phase flow measurement with high precision plays a significant role in safe and efficient operation of nuclear reactor. This paper focuses on the void fraction of gas-liquid flow in a non-transparent tube with the inner diameter of 13mm. X-ray computer tomography (CT) is considered to be effective for two-phase flow measurement because of its good qualities of non-invasion. However, due to the mechanical limitation of rotating speed, solutions for fast CT system to reduce the scanning time have been suggested relied on multiple sources and detectors. Recently, carbon nanotube (CNT) based X-ray source has been applied to CT imaging, significantly improves the temporal resolution by increasing the number of sources and avoid the gantry rotation. This paper proposes a potential static CT system design for the imaging of two-phase flow in straight steel tube. The setup of this system employed 90 couples of CNT X-ray sources and detector arrays arranged in a circle. Gas-liquid flow was simulated with different sizes of spheroidic bubbles randomly placed in the water inside the tube. To substitute for the flow moving, the z-axis of the simulator was added according to the flow velocity and exposure duration. Iterative image reconstruction was applied for the inverse problem of density distribution, and the reconstruction result of the experiment indicates the static CT system is useful to distinguish the flow pattern and measure the void faction distribution of gas-liquid flow.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129005630","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yunkang Feng, Lei Li, Yantao Nie, Xin Jiao, Jian Yi Li, Si Jia Meng, Y. Li
The plate-shaped fuel element has good heat transfer characteristics, high average power density of the core, and low temperature of the fuel core, which is beneficial to improve the power-to-volume ratio of the core and ensure the safety of the core. Therefore, plate fuels are widely used in compact reactors such as research reactors, integrated reactors, and high-flux reactors. at present, most thermal-hydraulic analysis programs, such as RELAP, RETRAN, THEATRE, are mostly developed for large-scale pressurized water reactors using rod-shaped fuels. It is suitable for narrow rectangular channel of plate type fuel core. Based on this, this paper developed a set of thermal-hydraulic constitutive relation models suitable for narrow rectangular channels, including: flow resistance coefficient calculation model, wall heat transfer Coefficient calculation model, CHF calculation model, etc. The thermal-hydraulic constitutive relational model library of rectangular channel of plate-shaped fuel element is developed by using C++ language. In this paper, the developed constitutive relation model is transplanted into the reactor thermal-hydraulic real-time simulation program, and the IAEA 10MW material test reactor (MTR) benchmark is used to verify the developed rectangular channel thermal-hydraulic constitutive relation model library. Simulation analysis is carried out for two typical accident conditions, reactive introduction (RIA) and loss of flow accident (LOFA) defined in the benchmark problem. correctness.
{"title":"Development and Verification of Thermal-Hydraulic Constitutive Model for Rectangular Channel","authors":"Yunkang Feng, Lei Li, Yantao Nie, Xin Jiao, Jian Yi Li, Si Jia Meng, Y. Li","doi":"10.1115/icone29-92795","DOIUrl":"https://doi.org/10.1115/icone29-92795","url":null,"abstract":"\u0000 The plate-shaped fuel element has good heat transfer characteristics, high average power density of the core, and low temperature of the fuel core, which is beneficial to improve the power-to-volume ratio of the core and ensure the safety of the core. Therefore, plate fuels are widely used in compact reactors such as research reactors, integrated reactors, and high-flux reactors. at present, most thermal-hydraulic analysis programs, such as RELAP, RETRAN, THEATRE, are mostly developed for large-scale pressurized water reactors using rod-shaped fuels. It is suitable for narrow rectangular channel of plate type fuel core. Based on this, this paper developed a set of thermal-hydraulic constitutive relation models suitable for narrow rectangular channels, including: flow resistance coefficient calculation model, wall heat transfer Coefficient calculation model, CHF calculation model, etc. The thermal-hydraulic constitutive relational model library of rectangular channel of plate-shaped fuel element is developed by using C++ language. In this paper, the developed constitutive relation model is transplanted into the reactor thermal-hydraulic real-time simulation program, and the IAEA 10MW material test reactor (MTR) benchmark is used to verify the developed rectangular channel thermal-hydraulic constitutive relation model library. Simulation analysis is carried out for two typical accident conditions, reactive introduction (RIA) and loss of flow accident (LOFA) defined in the benchmark problem. correctness.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128902988","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Hao Wu, Liangzhi Yu, F. Niu, J. Tu, Shengyao Jiang
In decay heat removal processes of the pebble-bed high temperature gas-cooled reactors, particle-scale radiative heat transfer between spheres is complicated for modeling and numerical simulations with traditional approaches. Artificial intelligence (AI) provides a new aspect to solve the dense granular dynamics. A machine learning model was developed for predicting the obstructed view factor between all possible pebble pairs in the large-scale nuclear pebble bed. The view factor dataset is obtained by random generation for sphere positions and thermal ray tracing method by CUDA paralleling for the view factor. The regression models are trained by gradient boosting decision tree (GBDT) method of XGBoost software for 2 ∼ 10 spheres cases. It is shown that the model performance will be greatly improved without overfitting by adding more trees rather than going deeper for every tree to reach R2 scores greater than 0.999. For engineering application, the trained XGboost models are applied to predict view factors in large-scale nuclear pebble bed during decay heat removal processes. From the transient numerical results, it takes about 10 h to get its maximum 1520°C only with thermal radiation and it is still less than the design upper limit.
{"title":"Machine Learning Modelling of Decay Heat Removal in High Temperature Gas-Cooled Reactor","authors":"Hao Wu, Liangzhi Yu, F. Niu, J. Tu, Shengyao Jiang","doi":"10.1115/icone29-92695","DOIUrl":"https://doi.org/10.1115/icone29-92695","url":null,"abstract":"\u0000 In decay heat removal processes of the pebble-bed high temperature gas-cooled reactors, particle-scale radiative heat transfer between spheres is complicated for modeling and numerical simulations with traditional approaches. Artificial intelligence (AI) provides a new aspect to solve the dense granular dynamics. A machine learning model was developed for predicting the obstructed view factor between all possible pebble pairs in the large-scale nuclear pebble bed. The view factor dataset is obtained by random generation for sphere positions and thermal ray tracing method by CUDA paralleling for the view factor. The regression models are trained by gradient boosting decision tree (GBDT) method of XGBoost software for 2 ∼ 10 spheres cases. It is shown that the model performance will be greatly improved without overfitting by adding more trees rather than going deeper for every tree to reach R2 scores greater than 0.999. For engineering application, the trained XGboost models are applied to predict view factors in large-scale nuclear pebble bed during decay heat removal processes. From the transient numerical results, it takes about 10 h to get its maximum 1520°C only with thermal radiation and it is still less than the design upper limit.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"18 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115511863","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Huiqiang Xu, Shoubao Dai, Naixun Sun, Lin Sun, Yanqing Wang
Thermal compressor, a similar device like steam driven jet pump, is usually used to keep the vacuum level of condensers in nuclear power plant. Meanwhile, the utilization of thermal compressor is an effective way to recover the heat of exhausted steam from turbines, improving the energy efficiency of secondary circuit in nuclear power plants. The extreme entrainment ration of thermal compressor is a significant parameter to describe its working performance. The present research conducts the theoretical derivation of the related physical process inside the thermal compressor and builds a theoretical model calculating the maximum entrainment ration. The results show that the present model can predict the ejection ability of the thermal compressor with satisfactory accuracy and application scope. The average relative deviation between calculated and experimental result is 9.7% and the maximum deviation does not exceed 23%. Moreover, the results of the model calculation indicate that the ejection ability is enhanced with the increase of primary and suction steam pressure, but weakened by the increase of compressed steam pressure. The entrainment ration increases with the superheat degree of primary steam because of the increasing critical velocity while the superheat degree of suction steam makes no obvious influence.
{"title":"Calculation Method and Analysis on Thermal Compressor Ejection Characteristics","authors":"Huiqiang Xu, Shoubao Dai, Naixun Sun, Lin Sun, Yanqing Wang","doi":"10.1115/icone29-92689","DOIUrl":"https://doi.org/10.1115/icone29-92689","url":null,"abstract":"\u0000 Thermal compressor, a similar device like steam driven jet pump, is usually used to keep the vacuum level of condensers in nuclear power plant. Meanwhile, the utilization of thermal compressor is an effective way to recover the heat of exhausted steam from turbines, improving the energy efficiency of secondary circuit in nuclear power plants. The extreme entrainment ration of thermal compressor is a significant parameter to describe its working performance. The present research conducts the theoretical derivation of the related physical process inside the thermal compressor and builds a theoretical model calculating the maximum entrainment ration. The results show that the present model can predict the ejection ability of the thermal compressor with satisfactory accuracy and application scope. The average relative deviation between calculated and experimental result is 9.7% and the maximum deviation does not exceed 23%. Moreover, the results of the model calculation indicate that the ejection ability is enhanced with the increase of primary and suction steam pressure, but weakened by the increase of compressed steam pressure. The entrainment ration increases with the superheat degree of primary steam because of the increasing critical velocity while the superheat degree of suction steam makes no obvious influence.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"56 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129478691","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Yi-quan Zhou, B. Kuang, Xin Wang, Shuting Wang, W. Hu, Lixia Ren
As one of the proposed six types of Gen-IV nuclear energy systems, lead-cooled fast reactor (LFR) has some advantages in safety, economy, sustainability, and proliferation prevention. With its low melting point, chemical inertia, high boiling point temperature, pretty good neutronics and γ shielding ability, lead bismuth eutectic (LEB) has been one of the common choices of lead-base coolant for fast reactors. A certain natural circulation capacity might be achieved in LBE flow systems due to its relatively high thermal expansion and thus induced buoyancy. Consequently, for the purpose of enhancing natural safety performance and operation economy, as well as for other specific needs, natural circulation is, in recent years, considered for the main heat transmission system or residual heat removal systems in some LBE fast reactor designs. Study of LBE natural circulation along with its heat transmission performance is thus of quite significance for LBE natural circulation fast reactor design and natural safety performance improvement. In this paper, steady-state flow and heat transmission characteristics and behaviors of LBE natural circulation, as well as those of other coolant media (sodium and water), are theoretically and comparatively studied. Meanwhile, based on both the steady and transient natural circulation experiments on the LBE natural circulation test facility, namely, LNC-SJTU facility, the applicability of the fast reactor system analysis code FRTAC for LBE natural circulation transient simulation is preliminarily validated. And with the prediction of this FRTAC code, furtherly, effects of operating conditions as well as the corresponding fluid thermophysical properties, structural and geometric parameters of the loop, frictional and local resistances on the LBE natural circulation performances are quantitatively investigated.
{"title":"A Combined Analytical, Experimental and Simulation Investigation on LBE Natural Circulation Flow and Heat Transmission","authors":"Yi-quan Zhou, B. Kuang, Xin Wang, Shuting Wang, W. Hu, Lixia Ren","doi":"10.1115/icone29-93584","DOIUrl":"https://doi.org/10.1115/icone29-93584","url":null,"abstract":"\u0000 As one of the proposed six types of Gen-IV nuclear energy systems, lead-cooled fast reactor (LFR) has some advantages in safety, economy, sustainability, and proliferation prevention. With its low melting point, chemical inertia, high boiling point temperature, pretty good neutronics and γ shielding ability, lead bismuth eutectic (LEB) has been one of the common choices of lead-base coolant for fast reactors. A certain natural circulation capacity might be achieved in LBE flow systems due to its relatively high thermal expansion and thus induced buoyancy. Consequently, for the purpose of enhancing natural safety performance and operation economy, as well as for other specific needs, natural circulation is, in recent years, considered for the main heat transmission system or residual heat removal systems in some LBE fast reactor designs. Study of LBE natural circulation along with its heat transmission performance is thus of quite significance for LBE natural circulation fast reactor design and natural safety performance improvement.\u0000 In this paper, steady-state flow and heat transmission characteristics and behaviors of LBE natural circulation, as well as those of other coolant media (sodium and water), are theoretically and comparatively studied. Meanwhile, based on both the steady and transient natural circulation experiments on the LBE natural circulation test facility, namely, LNC-SJTU facility, the applicability of the fast reactor system analysis code FRTAC for LBE natural circulation transient simulation is preliminarily validated. And with the prediction of this FRTAC code, furtherly, effects of operating conditions as well as the corresponding fluid thermophysical properties, structural and geometric parameters of the loop, frictional and local resistances on the LBE natural circulation performances are quantitatively investigated.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"8 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128728732","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The fission products released from the core fuel in severe accidents are mainly transported in the form of aerosols and exist in the gas phase in the containment. Once the containment fails, it will cause radioactive leakage. As an important mitigation measure for severe accidents, containment spray can effectively reduce the containment pressure and aerosol concentration, as well as the release of radioactivity to the environment. In order to analyze the thermal-hydraulic and aerosol behavior in the containment after accidents, a containment spray removal model was added to the integrated severe accident analysis code ISAA. The removal mechanism considers spray droplets washing aerosols by inertial impaction, interception collection, diffusion, thermophoresis, and diffusiophoresis. The containment spray removal model was coupled with the containment thermal-hydraulic module in ISAA to calculate the rate constant for spray removal. Benchmark experiments were selected for assessment of the improved code. The comparison shows that the results of the thermal-hydraulic response and aerosol mass distribution simulated by ISAA are consistent with the experimental data trends. The improved code can accurately simulate the thermal-hydraulic response and aerosol mass distribution during aerosol removal by spray droplets. Implementation of the containment spray removal model in ISAA satisfies the analysis of aerosol behavior in the containment.
{"title":"Development of Containment Spray Removal Model for Integrated Severe Accident Analysis Code ISAA","authors":"Jishen Li","doi":"10.1115/icone29-92721","DOIUrl":"https://doi.org/10.1115/icone29-92721","url":null,"abstract":"\u0000 The fission products released from the core fuel in severe accidents are mainly transported in the form of aerosols and exist in the gas phase in the containment. Once the containment fails, it will cause radioactive leakage. As an important mitigation measure for severe accidents, containment spray can effectively reduce the containment pressure and aerosol concentration, as well as the release of radioactivity to the environment. In order to analyze the thermal-hydraulic and aerosol behavior in the containment after accidents, a containment spray removal model was added to the integrated severe accident analysis code ISAA. The removal mechanism considers spray droplets washing aerosols by inertial impaction, interception collection, diffusion, thermophoresis, and diffusiophoresis. The containment spray removal model was coupled with the containment thermal-hydraulic module in ISAA to calculate the rate constant for spray removal. Benchmark experiments were selected for assessment of the improved code. The comparison shows that the results of the thermal-hydraulic response and aerosol mass distribution simulated by ISAA are consistent with the experimental data trends. The improved code can accurately simulate the thermal-hydraulic response and aerosol mass distribution during aerosol removal by spray droplets. Implementation of the containment spray removal model in ISAA satisfies the analysis of aerosol behavior in the containment.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"25 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127143403","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The tight lattice fuel assemblies (P/D < 1.1) have huge application potential in the new generation of the nuclear reactor systems, which possesses the advantage of higher power density and higher fuel conversion ratios, etc. Spacer grids can fix the fuel rod position and enhance the heat transfer performance. However, the two-phase flow characteristics downstream of spacer grids in the tight lattice are still not clarified. In this paper, a gas-liquid two-phase flow experiment study measuring the void fraction distribution downstream of spacer grids without mixing vanes (SGWMVs) in a double subchannels tight lattice bundle was conducted using a double-layer wire-mesh sensor. The experiment channel and the SGWMVs were both up-scaled (1:2.7) considering the difference between the experimental flow conditions and the prototype flow conditions. The phase distributions for a series of flow conditions (0.07 m/s < jg < 1.04 m/s, 0.93 m/s < jl < 1.86 m/s) were measured at Z/Dh = 115.81 of the bare rod channel as the flow data upstream of the spacer grid. Whereafter, the SGWMVs was installed at the position of Z/Dh = 145.92. The phase distributions 45mm downstream of SGWMVs were measured successively for the same flow conditions. Three types of flow patterns were obtained, including the bubbly flow, the cap-bubbly flow, and the slug flow. And the effects of SGWMVs on different flow patterns were described and analyzed. The newly obtained data established the reliability database contributing to the development of the computational fluid dynamics codes and the interfacial area transport equation.
{"title":"Measurement of the Two-Phase Flow Void Fraction Downstream Of Spacer Grids in Tight Lattice Bundles Using Wire-Mesh Sensor","authors":"Xu Yan, Yao Xiao, Hengwei Zhang","doi":"10.1115/icone29-92346","DOIUrl":"https://doi.org/10.1115/icone29-92346","url":null,"abstract":"\u0000 The tight lattice fuel assemblies (P/D < 1.1) have huge application potential in the new generation of the nuclear reactor systems, which possesses the advantage of higher power density and higher fuel conversion ratios, etc. Spacer grids can fix the fuel rod position and enhance the heat transfer performance. However, the two-phase flow characteristics downstream of spacer grids in the tight lattice are still not clarified. In this paper, a gas-liquid two-phase flow experiment study measuring the void fraction distribution downstream of spacer grids without mixing vanes (SGWMVs) in a double subchannels tight lattice bundle was conducted using a double-layer wire-mesh sensor. The experiment channel and the SGWMVs were both up-scaled (1:2.7) considering the difference between the experimental flow conditions and the prototype flow conditions. The phase distributions for a series of flow conditions (0.07 m/s < jg < 1.04 m/s, 0.93 m/s < jl < 1.86 m/s) were measured at Z/Dh = 115.81 of the bare rod channel as the flow data upstream of the spacer grid. Whereafter, the SGWMVs was installed at the position of Z/Dh = 145.92. The phase distributions 45mm downstream of SGWMVs were measured successively for the same flow conditions. Three types of flow patterns were obtained, including the bubbly flow, the cap-bubbly flow, and the slug flow. And the effects of SGWMVs on different flow patterns were described and analyzed. The newly obtained data established the reliability database contributing to the development of the computational fluid dynamics codes and the interfacial area transport equation.","PeriodicalId":325659,"journal":{"name":"Volume 7B: Thermal-Hydraulics and Safety Analysis","volume":"29 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"127453730","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：