Pub Date : 2024-11-06DOI: 10.1016/j.pnucene.2024.105525
Kepiao Li , Zhiyuan Wu , Kui Zhang , Wenxi Tian , Suizheng Qiu
Upper plenum entrainment phenomenon occurs in the automatic depressurization process during the Small Break Loss of Coolant Accident (SBLOCA) in reactor pressure vessel, which may result in reactor disaster. The upper plenum entrainment experiments with and without reactor internals were carried out with air-water and steam-water as working mediums on the Automatic Depressurization and Entrainment Test Loop for Upper plenum entrainment (ADETEL-U) which scaled after AP1000 nuclear reactor. The experimental phenomena were observed by visualization method and the reliable data were collected and analyzed. The results indicate that the entrainment rate will increase with the increase of gas flow rate under the same , and the entrainment rate will decrease significantly with the decrease of the mixed liquid level when the range of is low. The results confirm that a large number of liquid droplets will be deposited on the surface of the reactor internals, which greatly reduces the entrainment rate. Under the same conditions, the entrainment rate with the reactor internals is about 10% of that without the reactor internals. There is a huge discrepancy between the existing pool entrainment rate models and the experimental data, with the maximum deviation of 200 times. Based on the experimental results, new upper plenum entrainment models for near surface region and high gas flux region of momentum controlled region are proposed. The error decreases by orders of magnitude compared to existing models, which suggested that the new model can accurately predict upper plenum entrainment phenomenon in the pressure vessel.
{"title":"Experimental and theoretical research on upper plenum entrainment with air-water and steam-water","authors":"Kepiao Li , Zhiyuan Wu , Kui Zhang , Wenxi Tian , Suizheng Qiu","doi":"10.1016/j.pnucene.2024.105525","DOIUrl":"10.1016/j.pnucene.2024.105525","url":null,"abstract":"<div><div>Upper plenum entrainment phenomenon occurs in the automatic depressurization process during the Small Break Loss of Coolant Accident (SBLOCA) in reactor pressure vessel, which may result in reactor disaster. The upper plenum entrainment experiments with and without reactor internals were carried out with air-water and steam-water as working mediums on the Automatic Depressurization and Entrainment Test Loop for Upper plenum entrainment (ADETEL-U) which scaled after AP1000 nuclear reactor. The experimental phenomena were observed by visualization method and the reliable data were collected and analyzed. The results indicate that the entrainment rate will increase with the increase of gas flow rate under the same <span><math><mrow><msubsup><mi>h</mi><mi>g</mi><mo>∗</mo></msubsup></mrow></math></span>, and the entrainment rate will decrease significantly with the decrease of the mixed liquid level when the range of <span><math><mrow><msubsup><mi>h</mi><mi>g</mi><mo>∗</mo></msubsup></mrow></math></span> is low. The results confirm that a large number of liquid droplets will be deposited on the surface of the reactor internals, which greatly reduces the entrainment rate. Under the same conditions, the entrainment rate with the reactor internals is about 10% of that without the reactor internals. There is a huge discrepancy between the existing pool entrainment rate models and the experimental data, with the maximum deviation of 200 times. Based on the experimental results, new upper plenum entrainment models for near surface region and high gas flux region of momentum controlled region are proposed. The error decreases by orders of magnitude compared to existing models, which suggested that the new model can accurately predict upper plenum entrainment phenomenon in the pressure vessel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105525"},"PeriodicalIF":3.3,"publicationDate":"2024-11-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142592585","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-05DOI: 10.1016/j.pnucene.2024.105524
Cheng Peng , Chengfa Cao , Jiang Wu , Jian Deng
Liquid sodium can be treated as a prominent medium in many industrial fields, such as photovoltaic technology, chemical synthesis, nuclear industry, etc. However, it poses significant threats to the normal operation of related systems and facilities, and human life as well, due to its potential combustion risk, particularly when multi-leakages take place. Sodium spray combustion is the most severe one, in which spray dynamic process may intensify the heat transfer and subsequent combustion process. In this work, the applicability of the droplet break-up model is firstly confirmed using numerical simulations of liquid sodium spray by Fluent code, and the impact of spray interference on combustion kinetics is examined. The Euler-Lagrange approach, which accounts for droplet break-up, collision, and agglomeration during the spray combustion process, is used in the simulation. Three-dimensional simulations of liquid sodium spray fire are then conducted, in the light of two classical experiments all around the world. The simulated volume-mean air temperature shows an error margin of less than 4%. The thermodynamic characteristics of sodium spray fire in the situation of dual-jets is further investigated. The findings indicate that the spray interference has a greater impact on the sodium content threshold and the corresponding time at which the threshold can be achieved than temperature. When the nozzle spacing varies, the consequences of the spray interference on the droplets’ combustion change. The break-up impact outweighs the agglomeration effect when the nozzle spacing is larger, while the agglomeration effect is relatively stronger when the nozzle spacing is short. This conclusion can be appropriate under both low and high flow rate of liquid sodium. The present work can provide detailed information and mechanism on spray combustion under both single jet and dual-jets conditions, which is beneficial for the evaluation of the risk of real sodium spray fire in any closed environment.
{"title":"Comparison and analysis of combustion characteristics and interference effect between single burning sodium jet and the dual-jets","authors":"Cheng Peng , Chengfa Cao , Jiang Wu , Jian Deng","doi":"10.1016/j.pnucene.2024.105524","DOIUrl":"10.1016/j.pnucene.2024.105524","url":null,"abstract":"<div><div>Liquid sodium can be treated as a prominent medium in many industrial fields, such as photovoltaic technology, chemical synthesis, nuclear industry, etc. However, it poses significant threats to the normal operation of related systems and facilities, and human life as well, due to its potential combustion risk, particularly when multi-leakages take place. Sodium spray combustion is the most severe one, in which spray dynamic process may intensify the heat transfer and subsequent combustion process. In this work, the applicability of the droplet break-up model is firstly confirmed using numerical simulations of liquid sodium spray by Fluent code, and the impact of spray interference on combustion kinetics is examined. The Euler-Lagrange approach, which accounts for droplet break-up, collision, and agglomeration during the spray combustion process, is used in the simulation. Three-dimensional simulations of liquid sodium spray fire are then conducted, in the light of two classical experiments all around the world. The simulated volume-mean air temperature shows an error margin of less than 4%. The thermodynamic characteristics of sodium spray fire in the situation of dual-jets is further investigated. The findings indicate that the spray interference has a greater impact on the sodium content threshold and the corresponding time at which the threshold can be achieved than temperature. When the nozzle spacing varies, the consequences of the spray interference on the droplets’ combustion change. The break-up impact outweighs the agglomeration effect when the nozzle spacing is larger, while the agglomeration effect is relatively stronger when the nozzle spacing is short. This conclusion can be appropriate under both low and high flow rate of liquid sodium. The present work can provide detailed information and mechanism on spray combustion under both single jet and dual-jets conditions, which is beneficial for the evaluation of the risk of real sodium spray fire in any closed environment.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105524"},"PeriodicalIF":3.3,"publicationDate":"2024-11-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142586262","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-04DOI: 10.1016/j.pnucene.2024.105522
Xiuli Wang , Shenpeng Yang , YiFan Zhi , Wei Xu
Condition monitoring and identification are effective ways to ensure the safe and reliable operation of nuclear power pumps. However, the condition monitoring of the cutting impeller is blank. In order to effectively monitor and identify the operating status of nuclear power pump impellers corresponding to different cutting amounts. The paper collects the measured stator current signals of nuclear power pumps with 6 cutting quantities under 13 operating conditions. Variational Mode Decomposition (VMD) and Empirical Mode Decomposition (EMD) methods are utilized to analyze the state characteristics of the collected signals. The effect of different blade cutting amount on the signal characteristics of nuclear electric pump is obtained. The research results indicate as follow: when a single method is used to identify large impeller flow, the diagnostic accuracy of EMD and VMD can reach more than 90%, while Total Harmonic Distortion (THD) is less than 70%, and even less than 20% in some areas. However, for different impeller diameters and different flow rates, the identification accuracy of EMD and VMD is relatively low, only 60%. Under special working conditions, it can even be lower, with only about 50% at low flow rates between 0.2Q0-0.3Q0. EMD-VMD can accurately identify impellers of different diameters and different flow rates, and the accuracy of fault identification can be improved to over 90%, even higher than 95% in the range of 0.7Q0-1.2Q0. At the same time, the minimum flow rates of 0.2Q0 can also achieve 80% accuracy, which can effectively achieve fault diagnosis. The research results can provide data support for monitoring the operating status of self-cutting centrifugal pumps, which is of great significance for safe and stable operation.
{"title":"Research on impeller cutting of the nuclear pump based on MCSA","authors":"Xiuli Wang , Shenpeng Yang , YiFan Zhi , Wei Xu","doi":"10.1016/j.pnucene.2024.105522","DOIUrl":"10.1016/j.pnucene.2024.105522","url":null,"abstract":"<div><div>Condition monitoring and identification are effective ways to ensure the safe and reliable operation of nuclear power pumps. However, the condition monitoring of the cutting impeller is blank. In order to effectively monitor and identify the operating status of nuclear power pump impellers corresponding to different cutting amounts. The paper collects the measured stator current signals of nuclear power pumps with 6 cutting quantities under 13 operating conditions. Variational Mode Decomposition (VMD) and Empirical Mode Decomposition (EMD) methods are utilized to analyze the state characteristics of the collected signals. The effect of different blade cutting amount on the signal characteristics of nuclear electric pump is obtained. The research results indicate as follow: when a single method is used to identify large impeller flow, the diagnostic accuracy of EMD and VMD can reach more than 90%, while Total Harmonic Distortion (THD) is less than 70%, and even less than 20% in some areas. However, for different impeller diameters and different flow rates, the identification accuracy of EMD and VMD is relatively low, only 60%. Under special working conditions, it can even be lower, with only about 50% at low flow rates between 0.2Q<sub>0</sub>-0.3Q<sub>0</sub>. EMD-VMD can accurately identify impellers of different diameters and different flow rates, and the accuracy of fault identification can be improved to over 90%, even higher than 95% in the range of 0.7Q<sub>0</sub>-1.2Q<sub>0</sub>. At the same time, the minimum flow rates of 0.2Q<sub>0</sub> can also achieve 80% accuracy, which can effectively achieve fault diagnosis. The research results can provide data support for monitoring the operating status of self-cutting centrifugal pumps, which is of great significance for safe and stable operation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105522"},"PeriodicalIF":3.3,"publicationDate":"2024-11-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142578737","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-11-02DOI: 10.1016/j.pnucene.2024.105520
Timothy G. Lane, Shripad T. Revankar
A group of small nuclear reactors that are less than 20 MWe are often referred to as microreactors. This review provides recent advances in the nuclear reactor fuel and core design technology leading to compact microreactor designs, design features, types of microreactors currently considered in the industry and studied by researcher, regulatory design criteria, and deployment potentials for these new microreactors. This review indicates that there are a wide variety of microreactor designs being developed, some of which use coolant other than water such as liquid metal (e.g., sodium), helium gas or molten salt in order to achieve their operational objectives. Some of these designs utilize passive heat pipes in order to transfer heat from the reactor cores. Others make use of helium gas due to its compatibility at high temperature and inert nature. Currently there are no operating reactors which utilize either of these technologies for power generation. To aid the technology commercialization the nuclear regulatory bodies like US NRC are developing new design criteria and the licensing process to assess the microreactors for design certification, construction, and operation. The review indicates that there are potential design criteria challenges for the microreactors. For example, helium reactors need to show that the heat can be dispersed efficiently and passively, and the heat pipe reactors need to demonstrate that the coolant in their heat pipes will not escape the primary boundary. The US NRC has developed design criteria for microreactors are highlighted in the review.
{"title":"Advances in technology, design and deployment of microreactors- a review","authors":"Timothy G. Lane, Shripad T. Revankar","doi":"10.1016/j.pnucene.2024.105520","DOIUrl":"10.1016/j.pnucene.2024.105520","url":null,"abstract":"<div><div>A group of small nuclear reactors that are less than 20 MWe are often referred to as microreactors. This review provides recent advances in the nuclear reactor fuel and core design technology leading to compact microreactor designs, design features, types of microreactors currently considered in the industry and studied by researcher, regulatory design criteria, and deployment potentials for these new microreactors. This review indicates that there are a wide variety of microreactor designs being developed, some of which use coolant other than water such as liquid metal (e.g., sodium), helium gas or molten salt in order to achieve their operational objectives. Some of these designs utilize passive heat pipes in order to transfer heat from the reactor cores. Others make use of helium gas due to its compatibility at high temperature and inert nature. Currently there are no operating reactors which utilize either of these technologies for power generation. To aid the technology commercialization the nuclear regulatory bodies like US NRC are developing new design criteria and the licensing process to assess the microreactors for design certification, construction, and operation. The review indicates that there are potential design criteria challenges for the microreactors. For example, helium reactors need to show that the heat can be dispersed efficiently and passively, and the heat pipe reactors need to demonstrate that the coolant in their heat pipes will not escape the primary boundary. The US NRC has developed design criteria for microreactors are highlighted in the review.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105520"},"PeriodicalIF":3.3,"publicationDate":"2024-11-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142572360","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-30DOI: 10.1016/j.pnucene.2024.105468
Jinsuo Zhang, Nagihan Karakaya
Molten salt has broad applications in nuclear energy systems such as advanced nuclear reactors and pyroprocesses for spent fuel recycling. This paper discusses the available recycling processes for spent oxide fuels based on molten salt. The processes include oxide reduction, halidation, as well as directly electrorefining in molten fluoride and chloride salts. The paper also discusses the current research gaps for recycling processes.
{"title":"Molten salt technologies for recycling spent nuclear oxide fuel","authors":"Jinsuo Zhang, Nagihan Karakaya","doi":"10.1016/j.pnucene.2024.105468","DOIUrl":"10.1016/j.pnucene.2024.105468","url":null,"abstract":"<div><div>Molten salt has broad applications in nuclear energy systems such as advanced nuclear reactors and pyroprocesses for spent fuel recycling. This paper discusses the available recycling processes for spent oxide fuels based on molten salt. The processes include oxide reduction, halidation, as well as directly electrorefining in molten fluoride and chloride salts. The paper also discusses the current research gaps for recycling processes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105468"},"PeriodicalIF":3.3,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554029","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-30DOI: 10.1016/j.pnucene.2024.105492
Sukhdeep Singh , Nirvik Sen , V.P. Patel , D. Banerjee , Shaji Karunakaran , Sanjay Kumar
This paper deals with a cerium based redox chemical process which is used for the surface removal and decontamination of stainless steel components arising from the operation and decommissioning of nuclear facilities. The process involves the corrosion of top contaminated metal surface (∼10–50 μm) by highly oxidizing Ce(IV) ions, in order to remove the entrapped radionuclides. The reduced Ce(III) ions thus generated after the oxidation of metal surface are continuously oxidized back to Ce(IV) by ozone. In this work, a lumped parameter model, accounting for the reduction of Ce(IV) to Ce(III) by steel components and the simultaneous oxidation of Ce(III) to Ce(IV) by ozone in a static mixer, has been developed. The model is experimentally validated by carrying out a corrosion experiment at 5 L scale in ∼0.4 M solution of Ce(IV) in 4 M HNO3 at room temperature, by using non-radioactive AISI SS304L stainless steel components of different geometries viz. plate, pipe, elbow and T-Joint with a total surface area of 348 cm2. Using the validated model, design simulations of a pilot metal decontamination facility are carried out, to illustrate the effect of process parameters on the equilibrium Ce(IV) concentration in the loop, which is critical to component corrosion and decontamination. Simulation results show that, for a given initial cerium salt concentration, increasing the gas/liquid flow rate, ozone concentration, initial nitric acid concentration, and solution volume reduces the rate of fall of equilibrium Ce(IV) concentration with time. However, an increase in the temperature and surface area of the components enhances the rate of fall of equilibrium concentration of Ce(IV) with time. Additionally, reducing acidity of nitric acid has been found to limit the treatment time of components. Furthermore, the choice of a Ce(III) or Ce(IV) salt, as a source of cerium ions, has been shown to have no effect on the corrosion of metal components in a long run, when ozone regeneration of Ce(IV) is employed. Among the components of various geometries, relatively higher corrosion rates have been observed for the components with a curved geometry or a weld joint. SEM images of the welded and non-welded components show the occurrence of intergranular corrosion due to ozonated Ce(IV) solution, which is the likely mechanism for the removal of radionuclides from the metal surface.
本文论述了一种基于铈的氧化还原化学工艺，该工艺用于去除和净化核设施运行和退役过程中产生的不锈钢部件的表面。该过程包括用高度氧化的 Ce（IV）离子腐蚀受污染的金属表面（10-50 μm），以清除夹带的放射性核素。金属表面氧化后产生的还原 Ce（III）离子会被臭氧持续氧化回 Ce（IV）。在这项工作中，建立了一个集合参数模型，该模型考虑了钢成分将 Ce（IV）还原为 Ce（III）以及臭氧在静态混合器中将 Ce（III）氧化为 Ce（IV）的情况。使用不同几何形状的非放射性 AISI SS304L 不锈钢部件，即板材、管材、弯头和 T 型接头（总表面积为 348 cm2），在室温 4 M HNO3 中 0.4 M 的 Ce(IV) 溶液中进行 5 L 规模的腐蚀实验，对模型进行了实验验证。利用经过验证的模型，对试验性金属去污设施进行了设计模拟，以说明工艺参数对环路中平衡 Ce（IV）浓度的影响，该浓度对部件腐蚀和去污至关重要。模拟结果表明，对于给定的初始铈盐浓度，增加气体/液体流速、臭氧浓度、初始硝酸浓度和溶液体积可降低平衡铈（IV）浓度随时间的下降速度。然而，温度和组分表面积的增加会提高 Ce（IV）平衡浓度随时间的下降速度。此外，还发现硝酸的酸度降低会限制组分的处理时间。此外，选择 Ce（III）或 Ce（IV）盐作为铈离子源，在使用 Ce（IV）的臭氧再生时，长期来看对金属部件的腐蚀没有影响。在各种几何形状的部件中，具有弯曲几何形状或焊接接头的部件的腐蚀率相对较高。焊接和非焊接部件的扫描电子显微镜图像显示，臭氧化 Ce（IV）溶液会导致晶间腐蚀，这可能是从金属表面去除放射性核素的机制。
{"title":"Mathematical modeling of ozone assisted cerium redox process for the surface removal of stainless steel components using a static mixer as gas-liquid contactor","authors":"Sukhdeep Singh , Nirvik Sen , V.P. Patel , D. Banerjee , Shaji Karunakaran , Sanjay Kumar","doi":"10.1016/j.pnucene.2024.105492","DOIUrl":"10.1016/j.pnucene.2024.105492","url":null,"abstract":"<div><div>This paper deals with a cerium based redox chemical process which is used for the surface removal and decontamination of stainless steel components arising from the operation and decommissioning of nuclear facilities. The process involves the corrosion of top contaminated metal surface (∼10–50 μm) by highly oxidizing Ce(IV) ions, in order to remove the entrapped radionuclides. The reduced Ce(III) ions thus generated after the oxidation of metal surface are continuously oxidized back to Ce(IV) by ozone. In this work, a lumped parameter model, accounting for the reduction of Ce(IV) to Ce(III) by steel components and the simultaneous oxidation of Ce(III) to Ce(IV) by ozone in a static mixer, has been developed. The model is experimentally validated by carrying out a corrosion experiment at 5 L scale in ∼0.4 M solution of Ce(IV) in 4 M HNO<sub>3</sub> at room temperature, by using non-radioactive AISI SS304L stainless steel components of different geometries viz. plate, pipe, elbow and T-Joint with a total surface area of 348 cm<sup>2</sup>. Using the validated model, design simulations of a pilot metal decontamination facility are carried out, to illustrate the effect of process parameters on the equilibrium Ce(IV) concentration in the loop, which is critical to component corrosion and decontamination. Simulation results show that, for a given initial cerium salt concentration, increasing the gas/liquid flow rate, ozone concentration, initial nitric acid concentration, and solution volume reduces the rate of fall of equilibrium Ce(IV) concentration with time. However, an increase in the temperature and surface area of the components enhances the rate of fall of equilibrium concentration of Ce(IV) with time. Additionally, reducing acidity of nitric acid has been found to limit the treatment time of components. Furthermore, the choice of a Ce(III) or Ce(IV) salt, as a source of cerium ions, has been shown to have no effect on the corrosion of metal components in a long run, when ozone regeneration of Ce(IV) is employed. Among the components of various geometries, relatively higher corrosion rates have been observed for the components with a curved geometry or a weld joint. SEM images of the welded and non-welded components show the occurrence of intergranular corrosion due to ozonated Ce(IV) solution, which is the likely mechanism for the removal of radionuclides from the metal surface.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105492"},"PeriodicalIF":3.3,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554032","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-30DOI: 10.1016/j.pnucene.2024.105512
Mohamed Y.M. Mohsen , Shlash A. Luhaib , Nassar Alnassar , Omer A. Magzoub , Mohamed A.E. Abdel-Rahman , Mohammed Sallah , A. Abdelghafar Galahom
Reactor reactivity control materials play a crucial role in managing the stability and efficiency of nuclear reactors by regulating neutron flux and maintaining the desired reactivity levels throughout the reactor's operational cycle. This study explores the feasibility of using transuranic dioxide (TRUO₂) as reactivity control materials in pebble bed modular reactor 400 (PBMR-400) with thorium-based fuel. The TRU elements (Np, Pu, Am, and Cm) were extracted from spent uranium dioxide (UO₂) with a discharge burnup of 45 MWD/kgHM, following 30 years of cooling. The investigation covered four Th233UO2/TRUO2 mixtures, with ThO2 concentrations ranging from 75% to 95% and TRUO2 from 5% to 25%. This aims to determine the optimal composition that maximizes the TRUO2 concentration and minimizes ThO2 while preserving reactor performance in order to achieve the longest fuel cycle length with lower keff at the beginning of the fuel cycle (BOC) to avoid excess reactivity issues. Comprehensive neutronic analyses were conducted on these fuel mixtures, including burn-up, safety parameters, and flux and power distributions. The findings showed significant improvements in the PBMR-400's neutronic performance with the proposed fuel materials. From a safety, and economic standpoint, the optimal configuration was found to be 85% ThO2 and 15% TRUO2, as it provided the longest fuel cycle length with less excess reactivity at BOC and lower PPF.
{"title":"Investigating the possible advantages of using different concentrations of transuranic elements with thorium-uranium dioxide as a fuel for PBMR-400","authors":"Mohamed Y.M. Mohsen , Shlash A. Luhaib , Nassar Alnassar , Omer A. Magzoub , Mohamed A.E. Abdel-Rahman , Mohammed Sallah , A. Abdelghafar Galahom","doi":"10.1016/j.pnucene.2024.105512","DOIUrl":"10.1016/j.pnucene.2024.105512","url":null,"abstract":"<div><div>Reactor reactivity control materials play a crucial role in managing the stability and efficiency of nuclear reactors by regulating neutron flux and maintaining the desired reactivity levels throughout the reactor's operational cycle. This study explores the feasibility of using transuranic dioxide (TRUO₂) as reactivity control materials in pebble bed modular reactor 400 (PBMR-400) with thorium-based fuel. The TRU elements (Np, Pu, Am, and Cm) were extracted from spent uranium dioxide (UO₂) with a discharge burnup of 45 MWD/kgHM, following 30 years of cooling. The investigation covered four Th<sup>233</sup>UO<sub>2</sub>/TRUO<sub>2</sub> mixtures, with ThO<sub>2</sub> concentrations ranging from 75% to 95% and TRUO<sub>2</sub> from 5% to 25%. This aims to determine the optimal composition that maximizes the TRUO<sub>2</sub> concentration and minimizes ThO<sub>2</sub> while preserving reactor performance in order to achieve the longest fuel cycle length with lower k<sub>eff</sub> at the beginning of the fuel cycle (BOC) to avoid excess reactivity issues. Comprehensive neutronic analyses were conducted on these fuel mixtures, including burn-up, safety parameters, and flux and power distributions. The findings showed significant improvements in the PBMR-400's neutronic performance with the proposed fuel materials. From a safety, and economic standpoint, the optimal configuration was found to be 85% ThO<sub>2</sub> and 15% TRUO<sub>2</sub>, as it provided the longest fuel cycle length with less excess reactivity at BOC and lower PPF.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105512"},"PeriodicalIF":3.3,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554031","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-30DOI: 10.1016/j.pnucene.2024.105513
Qiaoqiao Fan , Menghong Xie , Zheng Lu , Di Jiang , Di Yao , Mengyan Song
To address the time-consuming and costly problems associated with the analysis methods primarily adopted at present (experimental test and numerical simulation), a simplified calculation method and algorithm are proposed. According to the theoretical solution of the thin-walled cylinder under inner pressure, the relationship between stress and strain is analyzed theoretically and the formula for calculating the mechanical response of the containment is derived and established with consideration of the material properties of containment's each component. Finally, the corresponding software, named as NCMC, is proposed. Three different containment types are selected for verification and these cases provide both numerical data and measured data. The comparison results demonstrate that NCMC can have good accuracy with exceptionally high computational efficiency, as well as it is highly flexible and has a significantly wider application scenario. NCMC is put forward to provide a relatively simplified method compared with traditional finite element (FE) software. This simplified calculation algorithm requires only a few seconds to a few minutes to obtain mechanical response results, which has significant advantages as FE software takes days to finish the same analysis. Hence, NCMC can have a promising application.
针对目前主要采用的分析方法（实验测试和数值模拟）耗时长、成本高的问题，提出了一种简化的计算方法和算法。根据内压下薄壁圆柱体的理论解，从理论上分析了应力和应变之间的关系，并结合安全壳各部件的材料特性，推导和建立了安全壳力学响应的计算公式。最后，提出了相应的软件，命名为 NCMC。我们选择了三种不同类型的安全壳进行验证，这些案例同时提供了数值数据和测量数据。对比结果表明，NCMC 不仅具有很高的计算效率，而且具有很高的精确度，同时还具有很高的灵活性和更广阔的应用前景。与传统的有限元（FE）软件相比，NCMC 提供了一种相对简化的方法。这种简化的计算算法只需要几秒到几分钟就能得到机械响应结果，这与 FE 软件需要几天才能完成同样的分析相比具有显著优势。因此，NCMC 具有广阔的应用前景。
{"title":"A simplified calculation method and algorithm for mechanical responses of nuclear containment","authors":"Qiaoqiao Fan , Menghong Xie , Zheng Lu , Di Jiang , Di Yao , Mengyan Song","doi":"10.1016/j.pnucene.2024.105513","DOIUrl":"10.1016/j.pnucene.2024.105513","url":null,"abstract":"<div><div>To address the time-consuming and costly problems associated with the analysis methods primarily adopted at present (experimental test and numerical simulation), a simplified calculation method and algorithm are proposed. According to the theoretical solution of the thin-walled cylinder under inner pressure, the relationship between stress and strain is analyzed theoretically and the formula for calculating the mechanical response of the containment is derived and established with consideration of the material properties of containment's each component. Finally, the corresponding software, named as NCMC, is proposed. Three different containment types are selected for verification and these cases provide both numerical data and measured data. The comparison results demonstrate that NCMC can have good accuracy with exceptionally high computational efficiency, as well as it is highly flexible and has a significantly wider application scenario. NCMC is put forward to provide a relatively simplified method compared with traditional finite element (FE) software. This simplified calculation algorithm requires only a few seconds to a few minutes to obtain mechanical response results, which has significant advantages as FE software takes days to finish the same analysis. Hence, NCMC can have a promising application.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105513"},"PeriodicalIF":3.3,"publicationDate":"2024-10-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142554030","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-28DOI: 10.1016/j.pnucene.2024.105515
Jinyi Cao , Jianchuang Sun , Xiangfei Meng , Wenchao Zhang , Jincheng Wang , Qian Li , Benan Cai , Weihua Cai
Helical petal-shaped fuel rods have the characteristics of increased heat transfer area and self-supporting positioning, which makes them have great potential for application in small modular reactors. By employing the multi-scale coupled numerical model, the flow and heat transfer characteristics of the petal-shaped bundle in hexagonal arrangement were obtained under natural circulation conditions. The results indicated that the spatial structure of flow channel exhibited a centrally symmetric distribution. Consistent flow and heat transfer behaviors were obtained at symmetric positions. Additionally, fluid viscosity exerted the most significant influence on flow resistance coefficient. Meanwhile, vortices that develop in the opposite direction resulted in flow losses, which induced variations in the resistance coefficient along the channel. Finally, the applicability of existing flow and heat transfer correlations was evaluated under natural circulation conditions. This study has provided significant theoretical guidance for engineering application of the petal-shaped fuel rods.
{"title":"Numerical investigation on flow and heat transfer of petal-shaped fuel bundle in hexagonal arrangement under natural circulation conditions","authors":"Jinyi Cao , Jianchuang Sun , Xiangfei Meng , Wenchao Zhang , Jincheng Wang , Qian Li , Benan Cai , Weihua Cai","doi":"10.1016/j.pnucene.2024.105515","DOIUrl":"10.1016/j.pnucene.2024.105515","url":null,"abstract":"<div><div>Helical petal-shaped fuel rods have the characteristics of increased heat transfer area and self-supporting positioning, which makes them have great potential for application in small modular reactors. By employing the multi-scale coupled numerical model, the flow and heat transfer characteristics of the petal-shaped bundle in hexagonal arrangement were obtained under natural circulation conditions. The results indicated that the spatial structure of flow channel exhibited a centrally symmetric distribution. Consistent flow and heat transfer behaviors were obtained at symmetric positions. Additionally, fluid viscosity exerted the most significant influence on flow resistance coefficient. Meanwhile, vortices that develop in the opposite direction resulted in flow losses, which induced variations in the resistance coefficient along the channel. Finally, the applicability of existing flow and heat transfer correlations was evaluated under natural circulation conditions. This study has provided significant theoretical guidance for engineering application of the petal-shaped fuel rods.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105515"},"PeriodicalIF":3.3,"publicationDate":"2024-10-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530082","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Pub Date : 2024-10-26DOI: 10.1016/j.pnucene.2024.105514
Zongyun Wu, Tiancai Liu, Yufeng Lyu, Chunqiu Guo, Lin Sun
The TOPAZ-II thermionic space reactor system, which was designed by the Soviet Union, is characterized by its high degree of nonlinearity and positive temperature reactivity feedback. The thermionic space reactor exhibits characteristics of high inertia and significant delay in controlling its electrical power and outlet temperature. The simple PID controller is difficult to achieve good performance. To carry out controller design for thermionic space reactor, the simulation platform for thermionic space reactor is developed based on coupling between reactor system thermal-hydraulic code RESYS and control system simulator in this study. After that, based on the cascade control strategy and gain-scheduling, the reactor thermal controller, electric power controller, outlet temperature controller applicable for the full power range is designed with transfer function model and frequency domain analysis method. To validate the nonlinear electric power controller performance, the continuous minor step disturbances, major step disturbances, and ramp variation of electric power setpoint is simulated. The performance of outlet temperature controller is verified with the simulation result of step and ramp variation of outlet temperature setpoint. Thereafter, the start-up process of the thermionic space reactor TOPAZ-II is simulated and analyzed. The simulation result reveals that the controller designed in this paper can overcome the nonlinearity of the thermionic space reactor system and has good performance throughout the entire power range. Compared to traditional simple PID controller, the cascade controller has better performance and can achieve good control performance even in situations where simple PID controllers cannot function properly.
{"title":"Non-linear cascade control with gain-scheduling and startup control strategy study for thermionic space reactor TOPAZ-II","authors":"Zongyun Wu, Tiancai Liu, Yufeng Lyu, Chunqiu Guo, Lin Sun","doi":"10.1016/j.pnucene.2024.105514","DOIUrl":"10.1016/j.pnucene.2024.105514","url":null,"abstract":"<div><div>The TOPAZ-II thermionic space reactor system, which was designed by the Soviet Union, is characterized by its high degree of nonlinearity and positive temperature reactivity feedback. The thermionic space reactor exhibits characteristics of high inertia and significant delay in controlling its electrical power and outlet temperature. The simple PID controller is difficult to achieve good performance. To carry out controller design for thermionic space reactor, the simulation platform for thermionic space reactor is developed based on coupling between reactor system thermal-hydraulic code RESYS and control system simulator in this study. After that, based on the cascade control strategy and gain-scheduling, the reactor thermal controller, electric power controller, outlet temperature controller applicable for the full power range is designed with transfer function model and frequency domain analysis method. To validate the nonlinear electric power controller performance, the continuous minor step disturbances, major step disturbances, and ramp variation of electric power setpoint is simulated. The performance of outlet temperature controller is verified with the simulation result of step and ramp variation of outlet temperature setpoint. Thereafter, the start-up process of the thermionic space reactor TOPAZ-II is simulated and analyzed. The simulation result reveals that the controller designed in this paper can overcome the nonlinearity of the thermionic space reactor system and has good performance throughout the entire power range. Compared to traditional simple PID controller, the cascade controller has better performance and can achieve good control performance even in situations where simple PID controllers cannot function properly.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"178 ","pages":"Article 105514"},"PeriodicalIF":3.3,"publicationDate":"2024-10-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142530081","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
扫码关注我们
求助内容：
应助结果提醒方式：
京公网安备 11010802042870号