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检索条件"机构=DiAME Institute for Advanced Research in Multiphase Flow and Energy Transfer"

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SMR Rod Bundle CHF-Characteristics and Challenges in Experimental Investigation 20

SMR Rod Bundle CHF-Characteristics and Challenges in Experim...

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20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Yang, Bao-Wen Liu, Aiguo Han, Bin DEQD Delta Energy Group No.413Venture Building NO.1 Changjiang Road Economic and Technology Development Zone Jiaozhou Shandong Qingdao266300 China DiAME Institute for Advanced Research in Multiphase Flow and Energy Transfer No.1 Huiying Road Jiaozhou Shandong Qingdao266300 China

ISBN: (纸本)9780894487934

Critical heat flux (CHF) is a threshold of significant importance that limits the available power during normal operating conditions of the light water reactor (LWR) and defines the limit for safety margins during accident *** recent years, the Small Modular Reactor (SMR) has gained its popularity *** to its small modular and special design operation conditions, such as forced, mixed, and natural circulations, the thermal-hydraulic behaviors of new fuel design types need further *** those, CHF is of the most ***, considering the integral design nature of most of light water based SMR, the thermal-hydraulic characteristics might be significantly different from conventional LWR designs, especially PWR *** SMR, the conditions involved range from forced to mixed convection and further to natural circulation and reflooding under accident *** ranges of interest for key thermal-hydraulic parameters are also expanded, for example, the pressure and flow ranges from high to low, while quality ranges from low to *** to high power peaking resulting from negative feedback associated with short bundle, especially in the Soluble-Boron Free (SBF) SMR design, for the excessive broad range of concerned thermal-hydraulic conditions, as well as the potential exit reflood quenching phenomena introduced by the large pool of water over the relatively short fuel assemblies in the integral SMR design, there are wide ranges of different CHF phenomena associated with SMR rod bundle *** phenomena span from the Non-uniform Heating Homogeneous Nucleation Model (NHNM) with CHF appearing prior to the occurrence of OSV (Onset of Significant Void), to subcooled boiling, Dry-out, and reflood cooling, with CHF mechanisms ranging from local to non-local *** wide spectrum of phenomena introduce numerous challenges for rod bundle CHF testing: *** Pressure Low flow (LPLF) conditions and the potential for flow inst

关键词： Light water reactors

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Study on the mixing vane grid effect on flow field and bubble distribution in a 2×1 subchannel by PIV and high speed camera 20

Study on the mixing vane grid effect on flow field and bubbl...

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20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Han, Bin Yang, Bao-Wen Zhu, Xiaoliang Liu, Aiguo Key Laboratory of Energy Thermal Conversion and Control Ministry of Education School of Energy and Environment Southeast University Nanjing210096 China DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No. 1 HuiYing Road Jiaozhou Shandong Qingdao266300 China

ISBN: (纸本)9780894487934

Mixing Vane Grid (MVG) is one of the most important components in the fuel assembly of a Light Water Reactor (LWR). In addition to providing the necessary mechanical support for the rod bundle, MVGs play a critical role in mixing the coolant water and even the thermal field. Mixing effects introduced by MVGs could make the reactor core temperature uniform and prevent hot spots in, improving the Critical Heat Flux(CHF), and improving the safety and economy of LWR. This is especially important for the relatively subcooled Pressurized Water Reactor (PWR). In order to obtain high resolution flow field under the effect of MVG, a 2×1 flow visualization rod bundle experiment was set up with Particle Image Velocity (PIV) and high-speed camera measurements for the development and validation of an effective tool for the evaluation of MVG performance. This study shows the vane effect on bubble distribution as viewed by High seed camera, demonstrating that the vane has different effects on bubble dynamics at various fractions of volume. The experimental data could be used to evaluate the grid performance and provide reliable data for the validation of Computational Fluid Dynamics (CFD) models. © 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.

关键词： Computational fluid dynamics

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Downard Reflood Quenching and the Effects of Power Distribution and Mixing Vane Spacer Grid Near the Exit of Fuel Assemblies in SMRs and Other LWR Reactors

Downard Reflood Quenching and the Effects of Power Distribut...

引用

2024 Advances in Thermal Hydraulics, ATH 2024

作者： Yang, Bao-Wen Silber, Jeremy Yang, Stephanie H. Lepore, Christopher Yang, Eric Han, Bin Liu, Aiguo Congers Ave. HaverstrawNY10927 United States CARP Associates USA LLC 57N Rt. 9W HaverstrawNY10927 United States DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No. 1 HuiYing Road Jiaozhou Shandong Qingdao266300 China

ISBN: (纸本)9780894482205

Study of the different thermal hydraulic phenomena that may occur in LWR type SMRs and other advanced reactors has been a growing field of interest in recent years. Recent experimental work has already shown that one of the phenomena that is of highest concern is the downward reflood quenching phenomenon near the exit of fuel bundles in SMRs and other LWR reactors [1, Yang]. This is because the phenomenon not only impacts the rod bundle CHF and induces flow instability, but it might also pose a great risk to the establishment of natural circulation, a vital safety feature for advanced reactors that rely on the inherently safe design principle. However, due to lack of testing under conditions where this phenomenon is most likely to occur, there is very limited data available to aid in prediction of how the transient event occurs and the factors that influence it. In order to further investigate the fundamental of these phenomena and develop comprehensive correlations to predict and mitigate the occurrence of such phenomena, a substantial amount of data would need to be collected at a wide range of test conditions, experimental setups, and test procedures, which are very likely to have a direct impact on the applicability of the test results. Furthermore, studies show that the downward quenching effect is heavily influenced by the power distribution and the mixing vane spacer grid design, each of which will be explored in this paper. Initial results from constant temperature and pressure testing indicates that reflooding occurrence can be assessed using a simple power and flow relationship with properly inclusion of pressure, temperature, void distribution., power profile, and spacer grid designs to account for more complicated reactor environments and operation conditions. The inclusion of a properly designed spacer grid can ultimately aid in preventing the downward reflood quenching phenomenon from occurring, or mitigating the effects to promote CHF while facilitati

关键词： Quenching

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Numerical Study on Subcooled Boiling and CHF Phenomenon in Eccentric Annular Channels

Numerical Study on Subcooled Boiling and CHF Phenomenon in E...

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14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, NUTHOS 2024

作者： Zhang, Yuanjie Han, Bin Zhu, Xiaoliang Yang, Bao-Wen Hong, Guangwei Xing, Tianyang Qi, Siwei Liu, Aiguo Liu, Shenghui Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education School of Energy and Environment Southeast University Nanjing210096 China DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer Qingdao266300 China

ISBN: (纸本)9780894482212

In the period of the increasing energy crisis, nuclear energy is of particularly importantance. It is critical to understand the core behaviors of fuel assembly and ultimately increase the CHF by accurately predicting flow and heat transfer phenomena inside the rod bundle. In the reactor core, the flow is quite complicated, including single phase flow, subcooled two-phase flow and Critical Heat Flux(CHF). The precise computational analysis of how to enhance the safety of nuclear reactors and thereby reduce costs becomes especially significant. Using computational Fluid Dynamics (CFD), it is possible to extract three-dimensional characteristics such as pressure, velocity, and temperature, providing theoretical guidance for nuclear reactor design and operation. By comparing Bartolomei's experimental data to the validated two-phase boiling CFD models, the CFD codes with RPI boiling models are applied to the internal and external flow phenomena of the pipe. The subcooled boiling two phase flow cases are compared. The void fraction and heated wall temperature distribution along the flow direction are discussed and compared under internal and external flow and a comparative analysis is conducted for CHF. It was found that the cold wall of the annular pipe has a significant effect on boiling and CHF. The study further extends the comparison to eccentric annular pipe configurations, exploring the performances of subcooled boiling and CHF. The results offer valuable insights into the understanding of the flow and heat transfer in the rod bundle of fuel assembly. © 2024 NUTHOS. All Rights Reserved.

关键词： Nuclear reactor accidents

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Application development and verification of lead cooling system safety analysis program based on RELAP5 MOD4.0 20

Application development and verification of lead cooling sys...

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20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Xiangyang, Wang Yang, Liu Xingkang, Su Qijian, Chen Lu, Zhang Xianwen, Li Bin, Han Aiguo, Liu Bao-Wen, Yang Long, Gu School of Nuclear Science and Technology Lanzhou University China Institute of Modern Physics Chinese Academy of Science China School of Nuclear Science and Technology University of Chinese Academy of Science China Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education School of Energy and Environment Southeast University Nanjing210096 China DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer Qingdao China

ISBN: (纸本)9780894487934

Lead-cooled natural circulation system is an important research content of lead-cooled fast reactor system. In order to obtain the thermal-hydraulic characteristics of lead-cooled natural circulation system, based on system safety analysis program RELAP5/MOD4.0, this study added thermal-physical relationship, flow resistance relationship and convection-heat transfer relationship for liquid lead coolant, and established RELAP5-Pb module which can be used for thermal-hydraulic analysis of lead system. The RELAP5-Pb program was verified preliminarily based on the data of thermal physical properties and heat transfer relations of lead based materials, and the results show that the added data of physical properties and constitutive relations are effective. Finally, the natural circulation experiment of 800 kW liquid lead was simulated based on RELAP5-Pb program, and the distribution characteristics of lead coolant flow rate, temperature and pressure were obtained. The extended lead cooling system analysis module RELAP5-Pb in this study can perform thermal-hydraulic analysis on lead-cooled fast reactor. © 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.

关键词： Natural convection

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Low Pressure Low flow Rod Bundle CHF and Its Challenges in SMR 20

Low Pressure Low Flow Rod Bundle CHF and Its Challenges in S...

引用

20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Yang, Bao-Wen Han, Bin Liu, Aiguo DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No.1 HuiYing Road Jiaozhou Shandong Qingdao266300 China Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education School of Energy and Environment Southeast University Nanjing210096 China DEGNY Delta Energy Group New York 57 Congers Ave. HaverstrawNY10927 United States

ISBN: (纸本)9780894487934

Critical Heat Flux (CHF) in SMR fuel bundle is one of the most important topics recently, with increasingly research focusing on CHF experimental ***, because most of rod bundle CHF data were obtained at High Pressure and High Temperature working conditions, it does not cover all the concerned parameters in the reactor core, especially the Low Pressure and Low flow CHF application for SMR and other advanced reactors, which are of great *** the Fukushima Accident and the advancement of the inherently safe design, especially for the SMR reactor, low pressure, low flow, and high-quality rod bundle data is becoming critically important for safety operation and AOO *** instability induced Low Pressure Low flow (LPLF) CHF is one of the most challenging conditions for SMR CHF *** to most local CHF events, low pressure low flow, flow instability induced CHF is an integral system phenomenon driven by the flow and pressure oscillation resulting from various thermal-hydraulic *** of the direct simulation heaters were designed and built for steady state tests, with rapid detection of temperature excursion at only certain designed locations with a high risk of local CHF *** the entire channel is boiling off under low pressure low flow conditions, not only can the integral CHF event occur at any unexpected location without thermocouple protection, the rapid pressure and flow transient can easily damage heater rods or sometime even the flow *** is rather risky to the expensive simulating heaters in running such rod bundle CHF tests subject to rapid flow instability *** USNRC acknowledged such challenges in their recent CBT Framework (2019).Due to the difficulties, very few rod bundle CHF data were obtained under 10 *** designs either used non-representative data from entirely different driven CHF mechanisms (such as high or medium pressure conditions) with a very subjective and non-conservative "p

关键词： Thermocouples

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Liquid Level Monitoring and Quenching Front Tracking for SMR Rod Bundle CHF tests under Low Pressure, Low flow, High Quality Conditions 20

Liquid Level Monitoring and Quenching Front Tracking for SMR...

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20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Yang, Bao-Wen Dempsey, Cameron Han, Bin Liu, Aiguo Yang, Eric Smith, Raymond DEGNY Delta Energy Group New York 57 Conger Ave. HaverstrawNY10927 United States DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No. 1 HuiYing Road Jiaozhou Shandong Qingdao266300 China CARP Associates USA LLC 57 N Rt. 9W HaverstrawNY10927 United States

ISBN: (纸本)9780894487934

In most Light Water Reactor (LWR) based Small Modular Reactors (SMRs), one of the critical factors that might affect Critical Heat Flux (CHF) and other thermal-hydraulic phenomena under low flow, low pressure, high-quality conditions is the existence of high water column over relatively short fuel core in majority of the integral design SMR reactor vessels. With the existence of this water column over the relatively short and high-power peaking fuel bundle, the potential of a reflood quenching effect is highly plausible under low pressure, low flow and high-quality conditions as the location of the CHF event will be much closer to the exit plenum, especially in the case of exit power peaking. The results from CFD modeling and flow visualization tests presented in this paper demonstrate the importance of liquid level as well as its impacts on rod bundle CHF. Such a critical factor has often been overlooked by the prevailing SMR designs as many of the rod bundle CHF tests performed for SMR, including those with the licensing data, have either neglected the effects of the exit quenching effect, or were performed with non-prototypical heater rod designs or surrogate fluid that may not include such exit quenching phenomena. In this paper, the phenomenon of exit quenching on SMR rod bundle CHF has been introduced with the development of the proprietary On-line, Real-time Liquid Level Monitoring System (OR_LLMS) and On-line, Real-time dynamic Quenching Front Tracking Systems (OR_QFTS), which allow accurate measurements of rod bundle CHF as a function of both liquid level and degree of downward quenching under various test conditions. The future application is also presented, which will aid in the development of rod bundle CHF correlations that account for and take advantage of different water levels over the reactor core. © 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.

关键词： Quenching

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Study on performance evaluation of 2×2 annular fuel with spacer grid in PWR 20

Study on performance evaluation of 2×2 annular fuel with sp...

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20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Wang, Sipeng Li, Zhuoyao Yang, Bao-Wen Department of Nuclear Science and Technology Nanjing University of Aeronautics and Astronautics JiangSu Nanjing211106 China Key Laboratory of Nuclear Technology Application and Radiation Protection in Astronautics Nanjing University of Aeronautics and Astronautics Ministry of Industry China DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No. 1 HuiYing Road Shandong Qingdao Jiaozhou266300 China

ISBN: (纸本)9780894487934

The performance of fuel elements is crucial to the safety and economy of a nuclear reactor, with the annular fuel offering the potential for improved safety and increased core power density compared to traditional rod fuel. However, the unique shape of annular fuel presents new challenges for the thermal hydraulic performance of the grid spacer. This study presents a 2×2 annular fuel bundle channel model with a spacer grid, proposes a performance evaluation factor for the positioning grid, and establishes a method and index for evaluating grid performance. Numerical simulations of the 2×2 annular fuel bundle channel under various working conditions are conducted, and the established evaluation factors are utilized to evaluate the thermal and hydraulic performance of the annular fuel bundle channel with spacer grid under different operating conditions. © 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.

关键词： Pressurized water reactors

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Numerical simulation of turbulent mixing in wire-wrapped fuel assembly of liquid lead-bismuth-cooled fast reactor 20

Numerical simulation of turbulent mixing in wire-wrapped fue...

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20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

作者： Wang, Sipeng Zhang, Qindong Yang, Bao-Wen Department of Nuclear Science and Technology Nanjing University of Aeronautics and Astronautics JiangSu Nanjing211106 China Key Laboratory of Nuclear Technology Application and Radiation Protection in Astronautics Nanjing University of Aeronautics and Astronautics Ministry of Industry No.1 HuiYing Road Jiaozhou Shandong Qingdao266300 China DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No.1 HuiYing Road Jiaozhou Shandong Qingdao266300 China

ISBN: (纸本)9780894487934

Due to its excellent economy and safety performance, the lead-bismuth-cooled fast reactor is one of the most popular candidates among the fourth-generation reactors, with a high application *** the design of most lead-bismuth-cooled reactors, the fuel rods are arranged compactly in a triangle, and positioned with each other by wrapped wire, which enhances the turbulent mixing inside the fuel *** is necessary to study the turbulent mixing behavior of liquid lead bismuth eutectic in wire-wrapped fuel *** this paper, a model of lead bismuth reactor with wire-wrapped 19 rod bundle assembly is *** the one hand, the influence of wire spacers on the flow and heat transfer between channels is *** the internal sub-channel, the wire sweeps through the flow channel in multiple directions, while sweeps in only one direction in the external ***, it is necessary to distinguish different types of flow channels considering mixing *** the other hand, the effect of wire axial pitch on turbulent mixing is *** study will provide a theoretical basis for accurate prediction of coolant flow and heat distribution in subchannel code, which is helpful to improve the safety and economy of the lead-bismuth cooled reactor. © 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.

关键词： Reactor refueling

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Investigation of the Impacts of Downward Reflood Quenching Phenomena on Rod Bundle CHF Measurements and on Safety Margin in LWR type SMRs

Investigation of the Impacts of Downward Reflood Quenching P...

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14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety, NUTHOS 2024

作者： Yang, Bao-Wen Yang, Stephanie H. Silber, Jeremy Yang, Eric Han, Bin Liu, Aiguo Congers Ave. HaverstrawNY10927 United States CARP Associates USA LLC 57N Rt. 9W HaverstrawNY10927 United States DEQD Institute for Advanced Research in Multiphase Flow and Energy Transfer No. 1 HuiYing Road Jiaozhou Shandong Qingdao266300 China

ISBN: (纸本)9780894482212

With recent developments and emergence of the light water reactor (LWR) type Small Module Reactor (SMR), there has been an increasing demand for evaluation of different thermal-hydraulic characteristics of these type of reactors. Considering the commonly adapted integral design feature in LWR type SMR designs, the thermal-hydraulic characteristics may be significantly different than conventional LWR designs, especially Pressurized Water Reactor (PWR) type SMRs. Additionally, it is well known that due to the inherently safe design principle, the establishment of Natural Circulation (NC) is one of the most critical safety features for all advanced reactors, including most of the LWR type SMRs. Experimental tests have shown that one of the greatest challenges to the establishment of NC under accident conditions is the downward quenching phenomena near the exit of the fuel core as well as the potential countercurrent flow limiting (CCFL) phenomena that might develop between rising void in the boiling hot channels and the downward traveling subcooled water in the cold surrounding channels (such as low power or guide tube or instrumentation channels). Integral SMRs, in particular, have an increased probability of these phenomena occurring due to the high upper plenum to reactor core length ratio. The large volume of unheated coolant in the upper plenum will be one of the key driving forces of the downward quenching phenomena and is one of the major differences between SMR and traditional PWR designs. The highly non-uniform power distributions, both axially and radially, tend to help promote such localized downward reflood quenching phenomena as well. This paper will provide clarification and verification of the reflooding phenomena with CFD modeling and recent experimental observations, while also exploring the impacts and potential risks related to the reflood quenching phenomena in the integral designed LWR type SMRs. To further investigate these impacts, first the diff

关键词： Pressurized water reactors

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