Nuclear power plants (NPPs) are widely used in the world. After three nuclear accidents, people propose higher of the safety and reliability on NPPs. Reactor coolant system (RCS) in the NPP directly affects whether the heat can be exported and radioactivity can be inclusive. It plays an important role of the NPPs safety. So, it is great significance of fault diagnosis for RCS in NPP. Although many scholar had carried out research on fault diagnosis of NPPs, different networks may lead to different results in a system. Therefore, this paper chooses a system and uses different neural networks (NN) for comparative analysis which can provide advice for follow-up research. In the paper, RCS has been analyzed and typical fault have been analyzed through PCTRAN simulator. On this basis, two kinds of NN combined with fuzzy systems: radial basis function (RBF) and back propagation (BP) are used for fault diagnosis and comparative analysis. Loss of coolant accident, single pump failure, loss of feed water are set for simulation experiment. Simulation experiment shows that BP network’s hidden layer nodes is less than RBF-NN, but iteration speed of BP network is faster; accuracy of fault diagnosis based on BP-NN is higher than RBF-NN; fuzzy-NN for fault diagnosis is faster than NN.
{"title":"Research on Fault Diagnosis of Reactor Coolant Accident in Nuclear Power Plant Based on Radial Basis Function and Fuzzy Neural Network","authors":"Pengpeng Sun, Yong Liu, Guohua Wu, Zhiyong Duan","doi":"10.1115/icone2020-16138","DOIUrl":"https://doi.org/10.1115/icone2020-16138","url":null,"abstract":"\u0000 Nuclear power plants (NPPs) are widely used in the world. After three nuclear accidents, people propose higher of the safety and reliability on NPPs. Reactor coolant system (RCS) in the NPP directly affects whether the heat can be exported and radioactivity can be inclusive. It plays an important role of the NPPs safety. So, it is great significance of fault diagnosis for RCS in NPP.\u0000 Although many scholar had carried out research on fault diagnosis of NPPs, different networks may lead to different results in a system. Therefore, this paper chooses a system and uses different neural networks (NN) for comparative analysis which can provide advice for follow-up research. In the paper, RCS has been analyzed and typical fault have been analyzed through PCTRAN simulator. On this basis, two kinds of NN combined with fuzzy systems: radial basis function (RBF) and back propagation (BP) are used for fault diagnosis and comparative analysis. Loss of coolant accident, single pump failure, loss of feed water are set for simulation experiment. Simulation experiment shows that BP network’s hidden layer nodes is less than RBF-NN, but iteration speed of BP network is faster; accuracy of fault diagnosis based on BP-NN is higher than RBF-NN; fuzzy-NN for fault diagnosis is faster than NN.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86020298","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The performance of the Reactor Core Isolation Cooling (RCIC) system under beyond design basis event (BDBE) conditions is not well-characterized. The operating band of the RCIC system is currently specified utilizing conservative assumptions, with restrictive operational guidelines not allowing for an adequate credit of the true capability of the system. For example, it is assumed that battery power is needed for RCIC operation to maintain the reactor pressure vessel (RPV) water level — a loss of battery power is conservatively assumed to result in failure of the RCIC turbopump system in a range of safety and risk assessments. However, the accidents at Fukushima Daiichi Nuclear Power Station (FDNPS) showed that the Unit 2 RCIC did not cease to operate following loss of battery power. In fact, it continued to inject water into the RPV for nearly 3 days following the earthquake. Improved understanding of Terry turbopump operations under BDBE conditions can support enhancement of accident management procedures and guidelines, promoting more robust severe accident prevention. Therefore, the U.S. Department of Energy (DOE), U.S. nuclear industry, and international stakeholders have funded the Terry Turbine Expanded Operating Band (TTEXOB) program. This program aims to better understand RCIC operations during BDBE conditions through combined experimental and modeling efforts. As part of the TTEXOB, airflow testing was performed at Texas A&M University (TAMU) of a small-scale ZS-1 and a full-scale GS-2 Terry turbine. This paper presents the corresponding efforts to model operation of the TAMU ZS-1 and GS-2 Terry turbines with Sandia National Laboratories’ (SNL) MELCOR code. The current MELCOR modeling approach represents the Terry turbine with a system of equations expressing the conservation of angular momentum. The joint analysis and experimental program identified that a) it is possible for the Terry turbine to develop the same power at different speeds, and b) turbine losses appear to be insensitive to the size of the turbine. As part of this program, further study of Terry turbine modeling unknowns and uncertainties is planned to support more extensive application of modeling and simulation to the enhancement of plant-specific operational and accident procedures.
{"title":"Computational Modeling of Terry Turbine Airflow Testing to Support the Expansion of Operating Band in Beyond Design Basis Conditions","authors":"L. Gilkey, N. Andrews, K. Ross, M. Solom","doi":"10.1115/icone2020-16547","DOIUrl":"https://doi.org/10.1115/icone2020-16547","url":null,"abstract":"\u0000 The performance of the Reactor Core Isolation Cooling (RCIC) system under beyond design basis event (BDBE) conditions is not well-characterized. The operating band of the RCIC system is currently specified utilizing conservative assumptions, with restrictive operational guidelines not allowing for an adequate credit of the true capability of the system. For example, it is assumed that battery power is needed for RCIC operation to maintain the reactor pressure vessel (RPV) water level — a loss of battery power is conservatively assumed to result in failure of the RCIC turbopump system in a range of safety and risk assessments. However, the accidents at Fukushima Daiichi Nuclear Power Station (FDNPS) showed that the Unit 2 RCIC did not cease to operate following loss of battery power. In fact, it continued to inject water into the RPV for nearly 3 days following the earthquake. Improved understanding of Terry turbopump operations under BDBE conditions can support enhancement of accident management procedures and guidelines, promoting more robust severe accident prevention. Therefore, the U.S. Department of Energy (DOE), U.S. nuclear industry, and international stakeholders have funded the Terry Turbine Expanded Operating Band (TTEXOB) program. This program aims to better understand RCIC operations during BDBE conditions through combined experimental and modeling efforts.\u0000 As part of the TTEXOB, airflow testing was performed at Texas A&M University (TAMU) of a small-scale ZS-1 and a full-scale GS-2 Terry turbine. This paper presents the corresponding efforts to model operation of the TAMU ZS-1 and GS-2 Terry turbines with Sandia National Laboratories’ (SNL) MELCOR code. The current MELCOR modeling approach represents the Terry turbine with a system of equations expressing the conservation of angular momentum. The joint analysis and experimental program identified that a) it is possible for the Terry turbine to develop the same power at different speeds, and b) turbine losses appear to be insensitive to the size of the turbine. As part of this program, further study of Terry turbine modeling unknowns and uncertainties is planned to support more extensive application of modeling and simulation to the enhancement of plant-specific operational and accident procedures.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81831075","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}
Modular high temperature gas-cooled reactor (MHTGR) is a small modular reactor (SMR) with inherent safety, which is suitable for load following to improve economic competitiveness. The heat regenerative system for MHTGR nuclear power plant, is crucial for the improvement of thermal efficiency. Traditionally, the enthalpy drop distribution method (EDM) is used to study the relationships between thermal efficiency and distribution of extraction steam. However, this strategy is mainly used for off-line design of steam turbine under rated conditions. For load following operation, it is hard to guarantee the extraction steam distribution of EDM due to the highly nonlinear “flowrate-pressure-temperature” coupling of the fluid network. Thus, in this paper, the thermal efficiency is derived analytically based on the steady state model of fluid network. Then the thermal efficiency optimization is cast into a nonlinear programming problem, in which physical constraints can be considered explicitly. The proposed method for extraction steam distribution is of significance for improving the thermal efficiency of normal operation of nuclear power plant.
{"title":"Thermal Efficiency Optimization of a Modular High Temperature Gas-Cooled Reactor Plant by Extraction Steam Distribution","authors":"Di Jiang, Z. Dong","doi":"10.1115/icone2020-16413","DOIUrl":"https://doi.org/10.1115/icone2020-16413","url":null,"abstract":"\u0000 Modular high temperature gas-cooled reactor (MHTGR) is a small modular reactor (SMR) with inherent safety, which is suitable for load following to improve economic competitiveness. The heat regenerative system for MHTGR nuclear power plant, is crucial for the improvement of thermal efficiency. Traditionally, the enthalpy drop distribution method (EDM) is used to study the relationships between thermal efficiency and distribution of extraction steam. However, this strategy is mainly used for off-line design of steam turbine under rated conditions. For load following operation, it is hard to guarantee the extraction steam distribution of EDM due to the highly nonlinear “flowrate-pressure-temperature” coupling of the fluid network. Thus, in this paper, the thermal efficiency is derived analytically based on the steady state model of fluid network. Then the thermal efficiency optimization is cast into a nonlinear programming problem, in which physical constraints can be considered explicitly. The proposed method for extraction steam distribution is of significance for improving the thermal efficiency of normal operation of nuclear power plant.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72964454","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}
In a hypothetical severe accident in a light water reactor (LWR) nuclear power plant, there is a possibility that molten core released from the reactor vessel gets in contact with water in the containment vessel. In this so-called fuel-coolant interactions (FCIs) process, the melt jet will breakup into fragments, which is one of the important factors for a steam explosion, as a potential threat to the integrity of the containment vessel. The particle method could directly and easily capture the large deformed interfaces by particle motions, benefiting from its Lagrangian description and meshless framework. In order to investigate the melt-jet breakup with solidification processes, a multiphase particle method with arbitrary high order scheme is presented in this study. In addition, an interfacial particle shifting scheme is developed to suppress the unnatural particle penetration between different phases. The convergence rate with different order is firstly confirmed by a verification test in terms of both explicit and implicit calculations. Then, a transient heat conduction between two materials is carried out and quite good results are obtained. After that, a rising bubble benchmark is performed to show the feasibility of modelling for deformation and collapse. Improvements of clear interface are indicated compared with previous reported results. Two important multiphase instabilities, namely the Rayleigh-Taylor instability and the Kelvin-Helmholtz instability, are studied since they play important roles during the melt-jet breakup. The results achieved so far indicate that the developed particle method is capable to analyze the melt-jet breakup with heat transfer.
{"title":"Development of a Multiphase Particle Method for Melt-Jet Breakup Behavior of Molten Core in Severe Accident","authors":"Zidi Wang, Y. Iwasawa, T. Sugiyama","doi":"10.1115/icone2020-16736","DOIUrl":"https://doi.org/10.1115/icone2020-16736","url":null,"abstract":"\u0000 In a hypothetical severe accident in a light water reactor (LWR) nuclear power plant, there is a possibility that molten core released from the reactor vessel gets in contact with water in the containment vessel. In this so-called fuel-coolant interactions (FCIs) process, the melt jet will breakup into fragments, which is one of the important factors for a steam explosion, as a potential threat to the integrity of the containment vessel. The particle method could directly and easily capture the large deformed interfaces by particle motions, benefiting from its Lagrangian description and meshless framework. In order to investigate the melt-jet breakup with solidification processes, a multiphase particle method with arbitrary high order scheme is presented in this study. In addition, an interfacial particle shifting scheme is developed to suppress the unnatural particle penetration between different phases. The convergence rate with different order is firstly confirmed by a verification test in terms of both explicit and implicit calculations. Then, a transient heat conduction between two materials is carried out and quite good results are obtained. After that, a rising bubble benchmark is performed to show the feasibility of modelling for deformation and collapse. Improvements of clear interface are indicated compared with previous reported results. Two important multiphase instabilities, namely the Rayleigh-Taylor instability and the Kelvin-Helmholtz instability, are studied since they play important roles during the melt-jet breakup. The results achieved so far indicate that the developed particle method is capable to analyze the melt-jet breakup with heat transfer.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85932721","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}
Byunghyun Choi, A. Nishida, K. Muramatsu, T. Itoi, T. Takada
After the 2011 Fukushima accident, the seismic regulations for nuclear power plants (NPP) in Japan have been strengthened to include countermeasures far beyond design-basis accidents. The importance of seismic probabilistic risk assessments, therefore, have been the focus of deserved attention. Generally, an uncertainty quantification has been a very important undertaking to assess for fragility in NPP buildings. Therefore, this study focuses on the reduction in epistemic uncertainty by aiming to clarify the seismic-response effects on NPP buildings based on different modeling methods. As a first step in this study, the authors compared the seismic-response effects using two modeling methods of analysis. To evaluate the seismic response, an analysis was performed on two building model types; these being the three-dimensional (3D) finite-element model and the sway-rocking model with a conventional lumped mass system. To input a ground motion, the authors adopted 200 types of simulated seismic ground motions, generated by fault-rupture models, using stochastic seismic source characteristics. For the uncertainty quantification, we conducted a statistical analysis of the seismic responses acquired from the two modeling methods based on the building response each ground-motion input, and quantitatively evaluated the uncertainty response by considering the different modeling methods. We found a clear difference in the modeling methods near the floor and wall openings. We also imparted our knowledge on these 3D effects for the seismic-response analysis.
{"title":"Uncertainty Quantification of Seismic Response of Reactor Building Considering Different Modeling Methods","authors":"Byunghyun Choi, A. Nishida, K. Muramatsu, T. Itoi, T. Takada","doi":"10.1115/icone2020-16862","DOIUrl":"https://doi.org/10.1115/icone2020-16862","url":null,"abstract":"\u0000 After the 2011 Fukushima accident, the seismic regulations for nuclear power plants (NPP) in Japan have been strengthened to include countermeasures far beyond design-basis accidents. The importance of seismic probabilistic risk assessments, therefore, have been the focus of deserved attention. Generally, an uncertainty quantification has been a very important undertaking to assess for fragility in NPP buildings. Therefore, this study focuses on the reduction in epistemic uncertainty by aiming to clarify the seismic-response effects on NPP buildings based on different modeling methods. As a first step in this study, the authors compared the seismic-response effects using two modeling methods of analysis. To evaluate the seismic response, an analysis was performed on two building model types; these being the three-dimensional (3D) finite-element model and the sway-rocking model with a conventional lumped mass system. To input a ground motion, the authors adopted 200 types of simulated seismic ground motions, generated by fault-rupture models, using stochastic seismic source characteristics. For the uncertainty quantification, we conducted a statistical analysis of the seismic responses acquired from the two modeling methods based on the building response each ground-motion input, and quantitatively evaluated the uncertainty response by considering the different modeling methods. We found a clear difference in the modeling methods near the floor and wall openings. We also imparted our knowledge on these 3D effects for the seismic-response analysis.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82234660","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}
Safety-related items in nuclear power plants are now generally placed separately from the non-safety-related items, but it was not strictly required before. Therefore, it is very important to study whether the non-safety-related items will affect the safety-related items when they are dropped down in an earthquake situation, which determines the safety of a nuclear power plant and its future life extension applications. This research was based on the cooling water system room with the safety and non-safety related items installed together, as an example to study whether the non-safety-related items such as vent pipes and DN50 fire fighting pipes arranged above will damage the DN300 pipes and valves arranged below when earthquakes occur. For the experiments, the relative positions of objects in the room was reproduced by 1: 1. The pressure-holding performance of the pipe was used as a criterion for the damage. The research results of the experiments show that when the 10-meter-long DN50 pipe was dropped from the position of 8 meters height and the 8-meter-long vent dropped from position of 3.6 meters height, they do not affect the integrity of the DN300 valve and pipe below. After the experiment, pressure drop in two hours for the pipe is less than 0.1%. The main body of the valve does not fail neither. The numerical simulation study also shows that there is no failure phenomenon in the simulation as well. Compared with the test results, the impact acceleration and the vent deformation both have the same trend.
{"title":"Study on the Interaction Between Safety-Related and Non Safety-Related Items in the Component Cooling Water System Room of the Qinshan Nuclear Power Plant in the Earthquake Condition","authors":"Liang Zhang, Gang Xu, Yue Wang, L. Chen, S. Zhou","doi":"10.1115/icone2020-16825","DOIUrl":"https://doi.org/10.1115/icone2020-16825","url":null,"abstract":"\u0000 Safety-related items in nuclear power plants are now generally placed separately from the non-safety-related items, but it was not strictly required before. Therefore, it is very important to study whether the non-safety-related items will affect the safety-related items when they are dropped down in an earthquake situation, which determines the safety of a nuclear power plant and its future life extension applications.\u0000 This research was based on the cooling water system room with the safety and non-safety related items installed together, as an example to study whether the non-safety-related items such as vent pipes and DN50 fire fighting pipes arranged above will damage the DN300 pipes and valves arranged below when earthquakes occur.\u0000 For the experiments, the relative positions of objects in the room was reproduced by 1: 1. The pressure-holding performance of the pipe was used as a criterion for the damage. The research results of the experiments show that when the 10-meter-long DN50 pipe was dropped from the position of 8 meters height and the 8-meter-long vent dropped from position of 3.6 meters height, they do not affect the integrity of the DN300 valve and pipe below. After the experiment, pressure drop in two hours for the pipe is less than 0.1%. The main body of the valve does not fail neither.\u0000 The numerical simulation study also shows that there is no failure phenomenon in the simulation as well. Compared with the test results, the impact acceleration and the vent deformation both have the same trend.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83818917","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}
T. Segawa, K. Kawaguchi, Katsunori Ishii, Masahiro Suzuki, Joji Tachihara, Kiyoto Takato, Takatoshi Okita, H. Satone, Michitaka Suzuki
To minimize retention of nuclear fuel materials in glove box components and curtail the external exposure dose, plutonium and uranium mixed oxide powder adhesion prevention technology involving nanoparticle coating of the acrylic panels of the glove box is developed. The nanoparticle coating reduces the van der Waals force between alumina particles and the acrylic test piece surface because of formation of nano-sized rugged surfaces. The nanoparticle coating reduces the minimum adhesion force normalized by the particle diameter between the uranium dioxide particle and the acrylic test piece surface, for the smallest particle of about 5 μm associated with desorption, this minimum adhesion force reduced to about 5%. The nanoparticle coating also lowers the adhered plutonium and uranium mixed oxide powder amounts on the acrylic test piece to about 10%. This study reveals that applying the nanoparticle coating to the acrylic panels of the glove box prevents adhesion of nuclear fuel materials. This method effectively reduces the retention of nuclear fuel materials in the glove box, lowers the external exposure dose, and improves the visibility of the acrylic panels.
{"title":"Preventing Nuclear Fuel Material Adhesion on Glove Box Components Using Nanoparticle Coating","authors":"T. Segawa, K. Kawaguchi, Katsunori Ishii, Masahiro Suzuki, Joji Tachihara, Kiyoto Takato, Takatoshi Okita, H. Satone, Michitaka Suzuki","doi":"10.1115/icone2020-16215","DOIUrl":"https://doi.org/10.1115/icone2020-16215","url":null,"abstract":"\u0000 To minimize retention of nuclear fuel materials in glove box components and curtail the external exposure dose, plutonium and uranium mixed oxide powder adhesion prevention technology involving nanoparticle coating of the acrylic panels of the glove box is developed. The nanoparticle coating reduces the van der Waals force between alumina particles and the acrylic test piece surface because of formation of nano-sized rugged surfaces. The nanoparticle coating reduces the minimum adhesion force normalized by the particle diameter between the uranium dioxide particle and the acrylic test piece surface, for the smallest particle of about 5 μm associated with desorption, this minimum adhesion force reduced to about 5%. The nanoparticle coating also lowers the adhered plutonium and uranium mixed oxide powder amounts on the acrylic test piece to about 10%. This study reveals that applying the nanoparticle coating to the acrylic panels of the glove box prevents adhesion of nuclear fuel materials. This method effectively reduces the retention of nuclear fuel materials in the glove box, lowers the external exposure dose, and improves the visibility of the acrylic panels.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78029062","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}
In decommissioning sodium-cooled fast reactors, the operators can be exposed to radiation during dismantling of cold trap equipment (C/T). The C/T is higher dose equipment because the C/T trapped tritium of fission products during the operation to purify the sodium coolant. In this study, thermal decomposition temperature and rate of sodium hydride (NaH) were measured as a fundamental research for development of “thermolysis” process prior to the dismantling. We measured the thermal decomposition temperature and rate using NaH powder (95.3%, Sigma-Aldrich) in alumina pan with ThermoGravimetry-Differential Thermal Analysis (TG-DTA) instrument (STA2500 Regulus, NETZSCH Japan). The heating rates of TG-DTA were set to β = 2.0, 5.0, 10.0 and 20.0 K/min. The DTA showed endothermic reaction and the TG showed two-steps mass-loss over 580K. This first-step mass-loss was consistent with change of chemical composition of the NaH with heating (NaH → Na+1/2H2). The thermal decomposition temperature and rate were obtained from the onset temperature of the mass-loss and the simplified Kissinger plots, respectively. Furthermore, we set to the thermal decomposition temperature of around 590K, and the mass-loss rates were measured. As a result, over 590K, the thermal decomposition occurred actively, and showed good agreement with the estimation curves obtained by the simplified Kissinger plots. The thermal decomposition rate strongly depended on the heating temperature.
{"title":"Measurement of Thermal Decomposition Temperature and Rate of Sodium Hydride","authors":"M. Kawaguchi","doi":"10.1115/icone2020-16423","DOIUrl":"https://doi.org/10.1115/icone2020-16423","url":null,"abstract":"\u0000 In decommissioning sodium-cooled fast reactors, the operators can be exposed to radiation during dismantling of cold trap equipment (C/T). The C/T is higher dose equipment because the C/T trapped tritium of fission products during the operation to purify the sodium coolant. In this study, thermal decomposition temperature and rate of sodium hydride (NaH) were measured as a fundamental research for development of “thermolysis” process prior to the dismantling.\u0000 We measured the thermal decomposition temperature and rate using NaH powder (95.3%, Sigma-Aldrich) in alumina pan with ThermoGravimetry-Differential Thermal Analysis (TG-DTA) instrument (STA2500 Regulus, NETZSCH Japan). The heating rates of TG-DTA were set to β = 2.0, 5.0, 10.0 and 20.0 K/min. The DTA showed endothermic reaction and the TG showed two-steps mass-loss over 580K. This first-step mass-loss was consistent with change of chemical composition of the NaH with heating (NaH → Na+1/2H2). The thermal decomposition temperature and rate were obtained from the onset temperature of the mass-loss and the simplified Kissinger plots, respectively. Furthermore, we set to the thermal decomposition temperature of around 590K, and the mass-loss rates were measured. As a result, over 590K, the thermal decomposition occurred actively, and showed good agreement with the estimation curves obtained by the simplified Kissinger plots. The thermal decomposition rate strongly depended on the heating temperature.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80758604","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}
In-Vessel Core Melt Retention (IVMR) strategy via external vessel cooling is widely applied for reactors of relatively low power like VVER-440. In this study, IVMR strategy was applied for Rivne-1, 2 units to prove the pressure vessel integrity. Based on initial data like heat flux for internal wall and external wall temperature, a series of calculations for different scenarios were performed. These calculations include non-elastic material properties: creep and plasticity. As the result, the wall ablation, radial displacements, stress and strains were obtained. To prove pressure vessel integrity four criterions have been checked. The first one is obvious — remaining wall thickness, to prove that that RPV won’t be melted right through. The second one is visco-plastic collapse — lack of monotonous increase in deformations, in case of FEM solution result convergence can be interpreted as resist against such failure. The third — sustainable external cooling, thus the gap between RPV (due to radial elongation) and thermal protection shield must be 10 mm at least. The last one is brittle strength, this calculation was performed on a separate model.
{"title":"In-Vessel Core Melt Retention Strategy Applied for the Rivne VVER-440 Unit","authors":"Y. Dubyk, V. Antonchenko","doi":"10.1115/icone2020-16913","DOIUrl":"https://doi.org/10.1115/icone2020-16913","url":null,"abstract":"\u0000 In-Vessel Core Melt Retention (IVMR) strategy via external vessel cooling is widely applied for reactors of relatively low power like VVER-440. In this study, IVMR strategy was applied for Rivne-1, 2 units to prove the pressure vessel integrity. Based on initial data like heat flux for internal wall and external wall temperature, a series of calculations for different scenarios were performed. These calculations include non-elastic material properties: creep and plasticity. As the result, the wall ablation, radial displacements, stress and strains were obtained. To prove pressure vessel integrity four criterions have been checked. The first one is obvious — remaining wall thickness, to prove that that RPV won’t be melted right through. The second one is visco-plastic collapse — lack of monotonous increase in deformations, in case of FEM solution result convergence can be interpreted as resist against such failure. The third — sustainable external cooling, thus the gap between RPV (due to radial elongation) and thermal protection shield must be 10 mm at least. The last one is brittle strength, this calculation was performed on a separate model.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85359985","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The Pressure Suppression Pool (PSP) in a Boiling Water Reactor (BWR) is served as a heat sink to prevent containment over-pressure. The steam can be injected through the multi-hole spargers. The development of thermal stratification where a thermocline with a large temperature gradient appears in the pool can lead to the higher pressure in the dry well compared with completely mixing pool conditions. Prediction of the thermal phenomenon in the pool is necessary for the support of system design and operation. Thus, the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed. The models can be applied to CFD code by using (i) source terms in the transport equations or (ii) using respective boundary conditions at the Steam Condensation Region (SCR). Previous validation against PPOOLEX and PANDA tests using the source terms approach faced challenges in momentum distribution. Therefore, a preliminary investigation of using the second method was performed. The encouraging results implied that it is possible to further develop this approach. The goal of this work is to further develop the EHS/EMS models for the steam injection through a multi-hole sparger through the SCR model (i.e approach (ii)) and to validate it against the experimental data obtained from PANDA HP5 tests. Modeling guidelines are proposed. The temperature evolutions and vertical velocity profiles of these tests are compared to the simulation results. The agreement suggests that this model can provide an adequate estimation of the pool behavior.
{"title":"Development of Effective Momentum Model for Steam Injection Through Multi-Hole Spargers Using a Condensation Region Approach","authors":"Xicheng Wang, D. Grishchenko, P. Kudinov","doi":"10.1115/icone2020-16852","DOIUrl":"https://doi.org/10.1115/icone2020-16852","url":null,"abstract":"\u0000 The Pressure Suppression Pool (PSP) in a Boiling Water Reactor (BWR) is served as a heat sink to prevent containment over-pressure. The steam can be injected through the multi-hole spargers. The development of thermal stratification where a thermocline with a large temperature gradient appears in the pool can lead to the higher pressure in the dry well compared with completely mixing pool conditions. Prediction of the thermal phenomenon in the pool is necessary for the support of system design and operation. Thus, the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models have been proposed. The models can be applied to CFD code by using (i) source terms in the transport equations or (ii) using respective boundary conditions at the Steam Condensation Region (SCR). Previous validation against PPOOLEX and PANDA tests using the source terms approach faced challenges in momentum distribution. Therefore, a preliminary investigation of using the second method was performed. The encouraging results implied that it is possible to further develop this approach. The goal of this work is to further develop the EHS/EMS models for the steam injection through a multi-hole sparger through the SCR model (i.e approach (ii)) and to validate it against the experimental data obtained from PANDA HP5 tests. Modeling guidelines are proposed. The temperature evolutions and vertical velocity profiles of these tests are compared to the simulation results. The agreement suggests that this model can provide an adequate estimation of the pool behavior.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77105435","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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