Kiyotaka Takito, O. Furuya, H. Kurabayashi, Kunio Sanpei
In Japan, most structures on the ground surface need seismic countermeasures because of frequently earthquakes. On the other hand, vibration isolation devices are applied to precision or important equipment in several facilities that dislikes vibration in order to reduce daily vibration. In general, vibration isolation devices are intended for high frequency and small amplitude range. However, it is difficult to cut off both vibration region caused by flying object collision and seismic motion with existing technologies. The authors propose insulation of equipment and vibration transmitted through the floor by floating equipment, and have. We have devised and built an air floating device that operates when a trigger input is applied to save the energy of this dynamically acting device. It was estimated by numerical calculation that the aero floating device keeps lifting stably in the condition with the air pressure in the auxiliary air chamber about 75 to 80 kPa. The performance specifications of the proposed device were verified from shaking table test. As a result, the effect of reducing the maximum acceleration by about 1/5 against the seismic motion of El Centro NS, Taft NS, Tohoku NS, and Hachinohe EW was confirmed by floating the mass on the frame assuming the equipment. From the obtained power spectrum diagram (PSD) of the response acceleration, it was confirmed that all frequency components up to 25 Hz is reduced by using proposed aero floating base isolation device.
{"title":"Study on Seismic Isolation and Hi-Frequency Vibration Isolation Technology for Equipment in Nuclear Power Plant Using Aero Floating Technique","authors":"Kiyotaka Takito, O. Furuya, H. Kurabayashi, Kunio Sanpei","doi":"10.1115/icone2020-16940","DOIUrl":"https://doi.org/10.1115/icone2020-16940","url":null,"abstract":"\u0000 In Japan, most structures on the ground surface need seismic countermeasures because of frequently earthquakes. On the other hand, vibration isolation devices are applied to precision or important equipment in several facilities that dislikes vibration in order to reduce daily vibration. In general, vibration isolation devices are intended for high frequency and small amplitude range. However, it is difficult to cut off both vibration region caused by flying object collision and seismic motion with existing technologies. The authors propose insulation of equipment and vibration transmitted through the floor by floating equipment, and have. We have devised and built an air floating device that operates when a trigger input is applied to save the energy of this dynamically acting device. It was estimated by numerical calculation that the aero floating device keeps lifting stably in the condition with the air pressure in the auxiliary air chamber about 75 to 80 kPa. The performance specifications of the proposed device were verified from shaking table test.\u0000 As a result, the effect of reducing the maximum acceleration by about 1/5 against the seismic motion of El Centro NS, Taft NS, Tohoku NS, and Hachinohe EW was confirmed by floating the mass on the frame assuming the equipment. From the obtained power spectrum diagram (PSD) of the response acceleration, it was confirmed that all frequency components up to 25 Hz is reduced by using proposed aero floating base isolation device.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"26 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114410119","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}
During a severe accident in a nuclear power plant, hydrogen would be generated due to the oxidation of metallic components in steam atmosphere. In the containment hydrogen would form a combustible mixture, posing a deflagration or even detonation risk threatening the integrity of the containment. In order to estimate possible loads generated by the hydrogen combustion, reliable numerical tools are needed to simulate the deflagration process. Recently, the French MITHYGENE project consortium and the European Technical Safety Organization Network (ETSON) organized a benchmark on hydrogen combustion to identify the current level of the computational tools in the area of hydrogen combustion simulation under a severe accident typical conditions. The benchmark was based on the experiments performed in the ENACCEF2 facility. This paper presents post-benchmark simulations of the selected ENACCEF2 facility premixed hydrogen combustion experiment. The presented simulations were performed using a custom-built turbulent combustion OpenFOAM solver based on the progress variable model. Turbulent flame acceleration phase in the acceleration tube was well predicted. Furthermore, the simulations were able to capture the interaction between the flame and shock wave which was generated by the turbulent deflagration flame and reflected at the end of the ENACCEF2 tube. The overall numerical results show good agreement with the qualitative and quantitative behavior of the velocity results and flame front propagation.
{"title":"Simulation of ENACCEF2 Premixed Hydrogen-Air Mixture Deflagration Experiment Using OpenFOAM","authors":"J. Jaseliūnaitė, Mantas Povilaitis","doi":"10.1115/icone2020-16241","DOIUrl":"https://doi.org/10.1115/icone2020-16241","url":null,"abstract":"\u0000 During a severe accident in a nuclear power plant, hydrogen would be generated due to the oxidation of metallic components in steam atmosphere. In the containment hydrogen would form a combustible mixture, posing a deflagration or even detonation risk threatening the integrity of the containment. In order to estimate possible loads generated by the hydrogen combustion, reliable numerical tools are needed to simulate the deflagration process. Recently, the French MITHYGENE project consortium and the European Technical Safety Organization Network (ETSON) organized a benchmark on hydrogen combustion to identify the current level of the computational tools in the area of hydrogen combustion simulation under a severe accident typical conditions. The benchmark was based on the experiments performed in the ENACCEF2 facility.\u0000 This paper presents post-benchmark simulations of the selected ENACCEF2 facility premixed hydrogen combustion experiment. The presented simulations were performed using a custom-built turbulent combustion OpenFOAM solver based on the progress variable model.\u0000 Turbulent flame acceleration phase in the acceleration tube was well predicted. Furthermore, the simulations were able to capture the interaction between the flame and shock wave which was generated by the turbulent deflagration flame and reflected at the end of the ENACCEF2 tube. The overall numerical results show good agreement with the qualitative and quantitative behavior of the velocity results and flame front propagation.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"26 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123799935","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}
High fidelity velocity field experimental data in a liquid metal plenum is presented and compared with numerical simulations. While work has already been established for fluids like air and water, research on low Pr fluids (Pr ≪ 1) (e.g. liquid metals) has fewer experimental data sets with validation-quality data. Work in advanced reactors using liquid metal coolant requires validated numerical simulations for safety analyses. The Gallium Thermal-hydraulic Experiment (GaTE) facility is outfitted with acoustic backscattering measurement techniques to generate the high fidelity distributed flow field data in a liquid metal plenum (a 1/20th scale of the Department of Energy’s sodium cooled Advanced Burner Test Reactor design). The high spatial and temporal resolution of the sensors are required to capture the fluctuations of velocity to allow a more direct comparison to the numerical simulations. For these simulations the coupled mass and momentum equations under the large eddy simulation (LES) framework were solved with the wall-adapting local eddy-viscosity (WALE) model for sub-grid scale formulations. Since the temperature transients of interest for reactor safety have a period of about a minute in the GaTE system, there may not be enough time to allow statistical tools to check one-to-one correspondence. So the data collection period for both data sets was extended to allow convergence of the mean and a larger sample size for other statistics during system steady-state, isothermal tests. Two characteristic velocities of the plenum inlet barrel were investigated (U = 40, 60 mm/s; Re = 7,000, 11,000). Probability distributions show good agreement between experiment and simulation with the difference only in the low-probability tails that LES is not expected to simulate. The time averaged mean axial distribution of the vertical velocity also shows good agreement between the two setups.
{"title":"Experimental Validation of CFD Models Capturing the Thermal-Hydraulics in Liquid Metal Cooled Reactor Plena","authors":"B. Ward, T. Hopkins, H. Bindra","doi":"10.1115/icone2020-16661","DOIUrl":"https://doi.org/10.1115/icone2020-16661","url":null,"abstract":"\u0000 High fidelity velocity field experimental data in a liquid metal plenum is presented and compared with numerical simulations. While work has already been established for fluids like air and water, research on low Pr fluids (Pr ≪ 1) (e.g. liquid metals) has fewer experimental data sets with validation-quality data. Work in advanced reactors using liquid metal coolant requires validated numerical simulations for safety analyses. The Gallium Thermal-hydraulic Experiment (GaTE) facility is outfitted with acoustic backscattering measurement techniques to generate the high fidelity distributed flow field data in a liquid metal plenum (a 1/20th scale of the Department of Energy’s sodium cooled Advanced Burner Test Reactor design). The high spatial and temporal resolution of the sensors are required to capture the fluctuations of velocity to allow a more direct comparison to the numerical simulations. For these simulations the coupled mass and momentum equations under the large eddy simulation (LES) framework were solved with the wall-adapting local eddy-viscosity (WALE) model for sub-grid scale formulations. Since the temperature transients of interest for reactor safety have a period of about a minute in the GaTE system, there may not be enough time to allow statistical tools to check one-to-one correspondence. So the data collection period for both data sets was extended to allow convergence of the mean and a larger sample size for other statistics during system steady-state, isothermal tests. Two characteristic velocities of the plenum inlet barrel were investigated (U = 40, 60 mm/s; Re = 7,000, 11,000). Probability distributions show good agreement between experiment and simulation with the difference only in the low-probability tails that LES is not expected to simulate. The time averaged mean axial distribution of the vertical velocity also shows good agreement between the two setups.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"20 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"129691625","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}
Fluoride salt-cooled High-temperature Reactor (FHR) is one of the advanced non-Light Water Reactor (non-LWR) designs, which adopts a low-pressure fluoride salt as the primary coolant, high working temperatures, coated-particle fuel, and a passive safety system for decay heat removal. However, tritium management is perceived as a critical issue for FHRs since tritium is a radiation hazard when inhaled or ingested and its production rate in FHRs is expected to be significantly higher compared to that in LWRs. To reduce FHR tritium release rates into the ambient, two tritium mitigation options, such as using Double-Wall Fluted-Tube Heat eXchangers (DWFT-HXs) with a tritium carrier or Single-Wall Fluted-Tube HXs (SWFT-HXs) with a tritium barrier, are therefore proposed for key HXs in FHRs, which potentially provide major pathways for tritium release due to their elevated temperatures and large surface areas. Tritium carriers investigated include gases, such as helium, and liquids, such as FLiBe, FLiNaK, and KF-ZrF4, while the tritium barrier investigated in this paper is silicon carbide (SiC) due to its low permeability for tritium. These proposed HX designs are then optimized, using a Non-dominated Sorting in Generic Algorithms (NSGA) optimization approach, for the Advanced High-Temperature Reactor (AHTR), one of the FHR designs with a large power output. A system-level mass transfer model is developed to evaluate the tritium transport in the two proposed design options for tritium mitigation in FHRs and quantitively analyze the tritium release/leakage rate from the reactor primary system. Our study shows that both the DWFT-HX design with helium as the tritium carrier and SWFT-HX design with SiC coating as the tritium barrier are able to reduce the total tritium leakage rate in FHRs to the same order of magnitude of the typical average tritium leakage rate in LWRs (1.9 Ci/day).
{"title":"Numerical Study of Tritium Mitigation Strategies for Fluoride Salt-Cooled High-Temperature Reactors","authors":"Sheng Zhang, Xiao Wu, Xiaodong Sun","doi":"10.1115/icone2020-16379","DOIUrl":"https://doi.org/10.1115/icone2020-16379","url":null,"abstract":"\u0000 Fluoride salt-cooled High-temperature Reactor (FHR) is one of the advanced non-Light Water Reactor (non-LWR) designs, which adopts a low-pressure fluoride salt as the primary coolant, high working temperatures, coated-particle fuel, and a passive safety system for decay heat removal. However, tritium management is perceived as a critical issue for FHRs since tritium is a radiation hazard when inhaled or ingested and its production rate in FHRs is expected to be significantly higher compared to that in LWRs. To reduce FHR tritium release rates into the ambient, two tritium mitigation options, such as using Double-Wall Fluted-Tube Heat eXchangers (DWFT-HXs) with a tritium carrier or Single-Wall Fluted-Tube HXs (SWFT-HXs) with a tritium barrier, are therefore proposed for key HXs in FHRs, which potentially provide major pathways for tritium release due to their elevated temperatures and large surface areas. Tritium carriers investigated include gases, such as helium, and liquids, such as FLiBe, FLiNaK, and KF-ZrF4, while the tritium barrier investigated in this paper is silicon carbide (SiC) due to its low permeability for tritium. These proposed HX designs are then optimized, using a Non-dominated Sorting in Generic Algorithms (NSGA) optimization approach, for the Advanced High-Temperature Reactor (AHTR), one of the FHR designs with a large power output.\u0000 A system-level mass transfer model is developed to evaluate the tritium transport in the two proposed design options for tritium mitigation in FHRs and quantitively analyze the tritium release/leakage rate from the reactor primary system. Our study shows that both the DWFT-HX design with helium as the tritium carrier and SWFT-HX design with SiC coating as the tritium barrier are able to reduce the total tritium leakage rate in FHRs to the same order of magnitude of the typical average tritium leakage rate in LWRs (1.9 Ci/day).","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"88 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126209080","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}
Buoyancy-driven flows are widespread in diverse engineering applications. Such flows have been studied in great detail theoretically, experimentally, and numerically. However, the fluid-dynamic instabilities and flow reversals of thermosiphon are still actively investigated. The presence of such instabilities limits the effectiveness of such devices for decay heat removal. Traditionally the stability analysis of natural convection loops has been confined to one-dimensional calculations, on the argument that the flow would be mono-dimensional when the ratio between the radius of the loop and the radius of the pipe is much larger than 1. Nevertheless, accurate velocity measurements of the flow in toroidal loops have shown that the flow presents three-dimensional effects. Previous works of the authors have shown that these structures can be seen in thermosiphons. In this paper, we aim to use modern CFD methods to investigate the three-dimensional flow in thermosiphons. This paper focuses on rectangular thermosiphons. In particular, we perform a series of high-fidelity simulations using the spectral element code Nek5000 to investigate the stability behavior of the flow in a rectangular thermosiphon. We compare the results with available existing experimental data from the L2 facility in Genoa. We examine in detail the flow structures generated. Moreover, in the past various authors have demonstrated that the overall behavior of the thermosiphon depends strongly on the boundary conditions (BCs). The simulation campaign was carried out with different BCs to investigate and confirm this effect. In particular, simulations with Dirichlet, Neumann and Robin BCs for heater and sink were performed.
{"title":"Computational Fluid Dynamics Simulation of a Single-Phase Rectangular Thermosiphon","authors":"Tri Nguyen, E. Merzari","doi":"10.1115/icone2020-16934","DOIUrl":"https://doi.org/10.1115/icone2020-16934","url":null,"abstract":"\u0000 Buoyancy-driven flows are widespread in diverse engineering applications. Such flows have been studied in great detail theoretically, experimentally, and numerically. However, the fluid-dynamic instabilities and flow reversals of thermosiphon are still actively investigated. The presence of such instabilities limits the effectiveness of such devices for decay heat removal.\u0000 Traditionally the stability analysis of natural convection loops has been confined to one-dimensional calculations, on the argument that the flow would be mono-dimensional when the ratio between the radius of the loop and the radius of the pipe is much larger than 1. Nevertheless, accurate velocity measurements of the flow in toroidal loops have shown that the flow presents three-dimensional effects. Previous works of the authors have shown that these structures can be seen in thermosiphons. In this paper, we aim to use modern CFD methods to investigate the three-dimensional flow in thermosiphons.\u0000 This paper focuses on rectangular thermosiphons. In particular, we perform a series of high-fidelity simulations using the spectral element code Nek5000 to investigate the stability behavior of the flow in a rectangular thermosiphon. We compare the results with available existing experimental data from the L2 facility in Genoa. We examine in detail the flow structures generated.\u0000 Moreover, in the past various authors have demonstrated that the overall behavior of the thermosiphon depends strongly on the boundary conditions (BCs). The simulation campaign was carried out with different BCs to investigate and confirm this effect. In particular, simulations with Dirichlet, Neumann and Robin BCs for heater and sink were performed.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"69 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123710063","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}
A. Uchibori, Mitsuhiro Aoyagi, T. Takata, H. Ohshima
The simulation system named SPECTRA for a severe accident in sodium-cooled fast reactors has been developed. The SPECTRA computes in- and ex-vessel phenomena and evaluates various scenarios during the severe accident. This paper provides a newly developed computational models for the ex-vessel phenomena including gas and aerosol transport, sodium-concrete interaction, and sodium fire as a part of the SPECTRA. The base module computing thermal hydraulics behavior by a lumped mass model was verified through the analysis of a 2-cells ventilation problem. The computational result of the SPECTRA agreed with the theoretical solutions both in the case with and without temperature change. The sodium-concrete interaction model was verified through code to code comparison. The computational result showed that ablation of a concrete surface started after surface temperature reached to a certain value. The computed ablation depth almost completely agreed with the result by the CONTAIN-LMR code. The ex-vessel module was applied to the computation assuming sodium leak from a reactor vessel and a primary cooling loop. This computation demonstrated increase of temperature and pressure due to sodium-concrete interaction and sodium fire.
{"title":"Development of Ex-Vessel Phenomena Analysis Model for Multi-Scenario Simulation System, SPECTRA","authors":"A. Uchibori, Mitsuhiro Aoyagi, T. Takata, H. Ohshima","doi":"10.1115/icone2020-16818","DOIUrl":"https://doi.org/10.1115/icone2020-16818","url":null,"abstract":"\u0000 The simulation system named SPECTRA for a severe accident in sodium-cooled fast reactors has been developed. The SPECTRA computes in- and ex-vessel phenomena and evaluates various scenarios during the severe accident. This paper provides a newly developed computational models for the ex-vessel phenomena including gas and aerosol transport, sodium-concrete interaction, and sodium fire as a part of the SPECTRA. The base module computing thermal hydraulics behavior by a lumped mass model was verified through the analysis of a 2-cells ventilation problem. The computational result of the SPECTRA agreed with the theoretical solutions both in the case with and without temperature change. The sodium-concrete interaction model was verified through code to code comparison. The computational result showed that ablation of a concrete surface started after surface temperature reached to a certain value. The computed ablation depth almost completely agreed with the result by the CONTAIN-LMR code. The ex-vessel module was applied to the computation assuming sodium leak from a reactor vessel and a primary cooling loop. This computation demonstrated increase of temperature and pressure due to sodium-concrete interaction and sodium fire.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"85 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116559085","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}
As key heat transfer system in small and medium size pressurized water reactors, once-through steam generators are important parts of energy exchange between primary and secondary circuits, and are very important for the design and operation of reactors. However, two-phase flow and heat transfer in once-through steam generators are very complicated. When a reactor experience power rising and descending transient, the heat removal of once-through steam generator, the flow rate, the inlet fluid temperature and outlet steam temperature will all change accordingly. Especially when a reactor is running at a low power, the flow rate of the secondary side of OTSG is extremely small and the single-phase region of the secondary side of OTSGs is also too small. The two-phase flow instability may occur, which has a serious impact on reactor operation and safety. So, a reasonable power-up and power-down transient scheme is required to ensure operational stability when starting up and shutting down a reactor. RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LCC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. Scholars all over the world have carried out a large number of analysis of two-phase flow stability using RELAP5, and the results are reliable. This paper takes once through steam generators with given structural parameters as the research object, and uses RELAP5 as the calculation tool. The influencing factors of flow instability are discussed in this paper, and the operating parameters of the fluid on the primary and secondary sides are designed to satisfy the flow stability under different powers. And a set of power-up and power-down schemes for stable operation is proposed.
蒸汽发生器作为中小型压水堆的关键传热系统，是一次回路和二次回路之间能量交换的重要部件，对反应堆的设计和运行具有十分重要的意义。然而，直通式蒸汽发生器的两相流动和传热是非常复杂的。当反应堆经历功率升降瞬态时，一次性蒸汽发生器的排热量、流量、进口流体温度和出口蒸汽温度都会发生相应的变化。特别是在反应器低功率运行时，OTSG二次侧流量极小，OTSG二次侧单相区也过小。可能出现两相流不稳定，严重影响反应堆的运行和安全。因此，为了保证反应堆在启动和关闭时的运行稳定性，需要合理的启动和关闭瞬态方案。RELAP5/MOD4.0是由Innovative System software, LCC开发的用于轻水反应堆(LWR)瞬态分析的商业软件。RELAP5经过多年的发展和完善，已经成为核电站各种模拟器分析计算的基础工具。国内外学者利用RELAP5进行了大量的两相流稳定性分析，结果是可靠的。本文以给定结构参数的一次过蒸汽发生器为研究对象，使用RELAP5作为计算工具。讨论了影响流动不稳定性的因素，设计了满足不同功率下流动稳定性的主、次侧流体运行参数。并提出了一套稳定运行的上电和下电方案。
{"title":"Power Rising and Descending Transient for the OTSG of a Small PWR","authors":"B. Jiang, Zhiwei Zhou, Z. Xia, Qian Sun","doi":"10.1115/icone2020-16294","DOIUrl":"https://doi.org/10.1115/icone2020-16294","url":null,"abstract":"\u0000 As key heat transfer system in small and medium size pressurized water reactors, once-through steam generators are important parts of energy exchange between primary and secondary circuits, and are very important for the design and operation of reactors. However, two-phase flow and heat transfer in once-through steam generators are very complicated. When a reactor experience power rising and descending transient, the heat removal of once-through steam generator, the flow rate, the inlet fluid temperature and outlet steam temperature will all change accordingly. Especially when a reactor is running at a low power, the flow rate of the secondary side of OTSG is extremely small and the single-phase region of the secondary side of OTSGs is also too small. The two-phase flow instability may occur, which has a serious impact on reactor operation and safety. So, a reasonable power-up and power-down transient scheme is required to ensure operational stability when starting up and shutting down a reactor.\u0000 RELAP5/MOD4.0 is a commercial software developed by Innovative System Software, LCC for transient analysis of light water reactors (LWR). After years of development and improvement, RELAP5 has been a basic tool for analysis and calculation of various simulators of nuclear power plants. Scholars all over the world have carried out a large number of analysis of two-phase flow stability using RELAP5, and the results are reliable.\u0000 This paper takes once through steam generators with given structural parameters as the research object, and uses RELAP5 as the calculation tool. The influencing factors of flow instability are discussed in this paper, and the operating parameters of the fluid on the primary and secondary sides are designed to satisfy the flow stability under different powers. And a set of power-up and power-down schemes for stable operation is proposed.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"46 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115801652","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}
Dynamic Uncertain Causality Graph (DUCG) is an innovative model developed recently on the basis of dynamic causality diagram (DCD) model, which has been proved to be reliable for fault diagnosis of nuclear power plants. DUCG can represent complex uncertain causal relationship graphically, with both high efficient inference and support of incomplete expression. Therefore, DUCG is often built much larger than Bayesian Network (BN). However, as the scale of real problem is so large, DUCG still has the problem of combination explosion. Stochastic Simulation is a common solution for it. However, it is almost impossible to use traditional sampling algorithms for DUCG because the joint probability of evidences could be less than 10−20. In this paper, the algorithm based on conditional stochastic simulation for the inference of DUCG was proposed. It obtains the probability of evidences by calculating the expectation of the conditional probability in sampling process instead of using the sampling frequency, which overcomes the difficulty. What’s more, this algorithm uses recursive reasoning method of DUCG to calculate conditional probability distributions of node for sampling, which means this process only depends on its parent nodes’ states. As a result, the algorithm features in lower time complexity. In addition, it has the potential of parallelization like other sampling algorithms. In conclusion, this algorithm is promising to provide a new solution to the inference of the DUCG in large-scale and complex state situations.
{"title":"Stochastic Simulation Method for Reasoning of Dynamic Uncertain Causality Graph (DUCG)","authors":"H. Nie, Qin Zhang","doi":"10.1115/icone2020-16327","DOIUrl":"https://doi.org/10.1115/icone2020-16327","url":null,"abstract":"\u0000 Dynamic Uncertain Causality Graph (DUCG) is an innovative model developed recently on the basis of dynamic causality diagram (DCD) model, which has been proved to be reliable for fault diagnosis of nuclear power plants. DUCG can represent complex uncertain causal relationship graphically, with both high efficient inference and support of incomplete expression. Therefore, DUCG is often built much larger than Bayesian Network (BN). However, as the scale of real problem is so large, DUCG still has the problem of combination explosion. Stochastic Simulation is a common solution for it. However, it is almost impossible to use traditional sampling algorithms for DUCG because the joint probability of evidences could be less than 10−20. In this paper, the algorithm based on conditional stochastic simulation for the inference of DUCG was proposed. It obtains the probability of evidences by calculating the expectation of the conditional probability in sampling process instead of using the sampling frequency, which overcomes the difficulty. What’s more, this algorithm uses recursive reasoning method of DUCG to calculate conditional probability distributions of node for sampling, which means this process only depends on its parent nodes’ states. As a result, the algorithm features in lower time complexity. In addition, it has the potential of parallelization like other sampling algorithms. In conclusion, this algorithm is promising to provide a new solution to the inference of the DUCG in large-scale and complex state situations.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131868586","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}
K. Zwijsen, H. Uitslag-Doolaard, F. Roelofs, J. Wallenius
SEALER (SwEdish Advanced Lead Reactor) is a passively safe lead-cooled reactor designed for commercial power production, under design by the LeadCold company. The reactor is modular in design, allowing for factory production and reduction in investment risk compared with new-build of large Light Water Reactors. Furthermore, its core is designed such that it can generate power for up to 25 years without the need of on-site fuel-cycle operations. The SEALER UK model has specifically been designed to produce base-load power on the UK grid. In the design and safety evaluation process, NRG is currently providing support to LeadCold Reactors with respect to thermal-hydraulic safety analyses utilizing Computational Fluid Dynamics (CFD) competences. The current paper gives a comprehensive description of a 3D CFD model created of SEALER UK Demo, which is a scaled-down demonstrator of SEALER UK. The geometry of the CFD model of SEALER UK Demo as well as the modelling approach and numerical settings are presented here. Assumptions were made in order to make it computationally feasible to perform simulations. These are discussed as well. Subsequently, the 3D CFD model is used to perform steady-state analyses of SEALER UK Demo operating under nominal conditions. Main parameters such as mass flow rates, temperatures and core pressure drops coming from the model match the design values well, with differences being at most a couple percent. Also, it is found that the margin to lead freezing with the current design parameters is more than 50K.
SEALER(瑞典先进铅反应堆)是一种被动安全的铅冷却反应堆，由LeadCold公司设计，用于商业发电。该反应堆采用模块化设计，与新建大型轻水反应堆相比，可实现工厂化生产，降低投资风险。此外，它的核心设计使得它可以在不需要现场燃料循环操作的情况下发电长达25年。SEALER英国模型专门设计用于在英国电网上产生基本负荷电力。在设计和安全评估过程中，NRG目前正在利用计算流体动力学(CFD)能力，为LeadCold反应堆提供热水力安全分析方面的支持。本文全面介绍了SEALER UK Demo创建的三维CFD模型，该模型是SEALER UK的缩小演示。本文介绍了SEALER UK Demo的CFD模型的几何形状、建模方法和数值设置。为了使模拟在计算上可行，进行了假设。也讨论了这些问题。随后，利用三维CFD模型对SEALER UK Demo在标称工况下的运行进行稳态分析。模型得到的质量流量、温度、堆芯压降等主要参数与设计值吻合较好，差异不超过几个百分点。同时发现，在当前设计参数下，导联冻结余量大于50K。
{"title":"Thermal-Hydraulic Design Support and Safety Analyses of SEALER UK Demo","authors":"K. Zwijsen, H. Uitslag-Doolaard, F. Roelofs, J. Wallenius","doi":"10.1115/icone2020-16526","DOIUrl":"https://doi.org/10.1115/icone2020-16526","url":null,"abstract":"\u0000 SEALER (SwEdish Advanced Lead Reactor) is a passively safe lead-cooled reactor designed for commercial power production, under design by the LeadCold company. The reactor is modular in design, allowing for factory production and reduction in investment risk compared with new-build of large Light Water Reactors. Furthermore, its core is designed such that it can generate power for up to 25 years without the need of on-site fuel-cycle operations. The SEALER UK model has specifically been designed to produce base-load power on the UK grid. In the design and safety evaluation process, NRG is currently providing support to LeadCold Reactors with respect to thermal-hydraulic safety analyses utilizing Computational Fluid Dynamics (CFD) competences. The current paper gives a comprehensive description of a 3D CFD model created of SEALER UK Demo, which is a scaled-down demonstrator of SEALER UK. The geometry of the CFD model of SEALER UK Demo as well as the modelling approach and numerical settings are presented here. Assumptions were made in order to make it computationally feasible to perform simulations. These are discussed as well. Subsequently, the 3D CFD model is used to perform steady-state analyses of SEALER UK Demo operating under nominal conditions. Main parameters such as mass flow rates, temperatures and core pressure drops coming from the model match the design values well, with differences being at most a couple percent. Also, it is found that the margin to lead freezing with the current design parameters is more than 50K.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"32 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115271073","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}
Zefeng Wang, Jian Deng, Libo Qian, R. Cai, Jinbiao Xiong, Lei Zhong, Yugao Ma
Quenching is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident (LOCA) in a nuclear reactor. In the present study, an experimental apparatus is designed and constructed with the purpose of conducting high-temperature transient pool boiling quenching experiments for zirconium (Zr-4) cylindrical test samples. Three thermocouples are inserted in the test sample to investigate the effect of axial distance on the minimum film boiling temperature. The Zr-4 rodlet is heated up to a temperature well above the minimum film boiling temperature (up to 600°C), and then plunged vertically in a quiescent pool of subcooled water. A data acquisition system is used to record the temperature of the embedded thermocouples with time. Data reduction is performed by an inverse heat conduction code to calculate the surface temperature and corresponding surface heat flux. A visualization study with a high-speed camera is conducted to record the quenching behavior on the test sample. It is found that the minimum film boiling temperature decreases with the axial distance, while the CHF temperature is relatively insensitive to the axial distance. The film boiling heat transfer coefficient decreases with surface temperature, and seems to be independent of axial distance. The quench front is observed to originate from the bottom and move upwards. It is found that the quench front velocity remains nearly constant in the lower region of the test sample, and significantly increases in the upper region.
{"title":"Experimental Investigation of the Transient Pool Boiling Heat Transfer on the Quenching of Vertical Rodlet in Water","authors":"Zefeng Wang, Jian Deng, Libo Qian, R. Cai, Jinbiao Xiong, Lei Zhong, Yugao Ma","doi":"10.1115/icone2020-16709","DOIUrl":"https://doi.org/10.1115/icone2020-16709","url":null,"abstract":"\u0000 Quenching is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident (LOCA) in a nuclear reactor. In the present study, an experimental apparatus is designed and constructed with the purpose of conducting high-temperature transient pool boiling quenching experiments for zirconium (Zr-4) cylindrical test samples. Three thermocouples are inserted in the test sample to investigate the effect of axial distance on the minimum film boiling temperature. The Zr-4 rodlet is heated up to a temperature well above the minimum film boiling temperature (up to 600°C), and then plunged vertically in a quiescent pool of subcooled water. A data acquisition system is used to record the temperature of the embedded thermocouples with time. Data reduction is performed by an inverse heat conduction code to calculate the surface temperature and corresponding surface heat flux. A visualization study with a high-speed camera is conducted to record the quenching behavior on the test sample. It is found that the minimum film boiling temperature decreases with the axial distance, while the CHF temperature is relatively insensitive to the axial distance. The film boiling heat transfer coefficient decreases with surface temperature, and seems to be independent of axial distance. The quench front is observed to originate from the bottom and move upwards. It is found that the quench front velocity remains nearly constant in the lower region of the test sample, and significantly increases in the upper region.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"49 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"121656344","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}
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