The pressurizer is responsible for controlling the pressure and temperature in the primary circuit of the nuclear power plant. It is basically a pressure vessel, filled partially with water and partially with steam at saturation state. The controlling process can be described by two regimes, either by the self-control regime or by the automatic control regime. This paper is describing the simplified automatic control regime on the model of the pressurizer in the primary circuit of VVER 1000 made in Dymola software with the ClaRa library. Reactor power change and corresponding steam generator power change are the actuators in the simulation. The behaviour of the coolant level and pressure in the pressurizer is simulated in the model and it is then compared with data provided by the supplier of VVER 1000.
{"title":"VVER 1000 Pressurizer System and Control Modelling in Dymola","authors":"Anna Fortová, Filip Ježek","doi":"10.1115/ICONE26-81263","DOIUrl":"https://doi.org/10.1115/ICONE26-81263","url":null,"abstract":"The pressurizer is responsible for controlling the pressure and temperature in the primary circuit of the nuclear power plant. It is basically a pressure vessel, filled partially with water and partially with steam at saturation state. The controlling process can be described by two regimes, either by the self-control regime or by the automatic control regime. This paper is describing the simplified automatic control regime on the model of the pressurizer in the primary circuit of VVER 1000 made in Dymola software with the ClaRa library. Reactor power change and corresponding steam generator power change are the actuators in the simulation. The behaviour of the coolant level and pressure in the pressurizer is simulated in the model and it is then compared with data provided by the supplier of VVER 1000.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"43 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116965884","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 best estimate plus uncertainty (BEPU) method is recommended by IAEA for nuclear safety analysis. Most of the existing BEPU methodologies rely on the uncertainty propagation of input parameters, while uncertainties of the constitutive models in best estimate codes tend not to be valued or even neglected. A structural method is proposed in this paper to quantify the uncertainties of the constitutive models. Different constitutive models will be classified according to the characteristics and corresponding method could be utilized for each model based on the method. Specific uncertainty quantification (UQ) methods adopted in this paper include the non-parametric curve estimation method, inverse method and design of experiment (DOE) method, and a model selection technique is adopted to opt the optimal model among all alternative models. The structural method is applied to the uncertainty evaluation of LOFT LP-02-6 experiment. Important models are identified, uncertainties of these models are quantified and propagated to the peak cladding temperature (PCT) through code calculations. Uncertainty of the PCT is quantified and the result shows that the calculated values could well envelop the experimental value.
{"title":"Investigation on Methods for Uncertainty Quantification of Constitutive Models and the Application in BEPU","authors":"Qingwen Xiong, J. Gou, J. Shan","doi":"10.1115/ICONE26-81425","DOIUrl":"https://doi.org/10.1115/ICONE26-81425","url":null,"abstract":"The best estimate plus uncertainty (BEPU) method is recommended by IAEA for nuclear safety analysis. Most of the existing BEPU methodologies rely on the uncertainty propagation of input parameters, while uncertainties of the constitutive models in best estimate codes tend not to be valued or even neglected. A structural method is proposed in this paper to quantify the uncertainties of the constitutive models. Different constitutive models will be classified according to the characteristics and corresponding method could be utilized for each model based on the method. Specific uncertainty quantification (UQ) methods adopted in this paper include the non-parametric curve estimation method, inverse method and design of experiment (DOE) method, and a model selection technique is adopted to opt the optimal model among all alternative models. The structural method is applied to the uncertainty evaluation of LOFT LP-02-6 experiment. Important models are identified, uncertainties of these models are quantified and propagated to the peak cladding temperature (PCT) through code calculations. Uncertainty of the PCT is quantified and the result shows that the calculated values could well envelop the experimental value.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"123452577","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}
Huijun Liang, K. Ge, Yapei Zhang, G. Su, W. Tian, S. Qiu
Hypothetical Core Disruptive Accidents (HCDA) are dominantly concerned during safety assessment and evaluation in Sodium-cooled Fast Reactor (SFR). With molten core materials discharged into liquid sodium, positive reactivity is potentially introduced due to sodium boiling and molten core compaction, which can cause terrible recriticality. The possibility of recriticality and efficient cooling on the relocated debris bed are significantly affected by the fragmentation behavior of molten core in liquid sodium. With few available mechanism models and benchmarks, many investigations have been conducted on the fragmentation characteristics during molten fuel-coolant interaction (MFCI). In the present study, molten copper is used for molten simulant to be discharged into the liquid sodium pool through guiding tube based on a multifunctional experimental facility (COSA). The simulants are heated by electromagnetic induction system in customized ceramic crucible and the molten materials are controlled by magnetic lifting system to be drained through the guiding tube into the bottom liquid sodium pool. Temperature variation and pressure change in the liquid sodium pool are acquired against the energy release during MFCI. Furthermore, the fragments cleaned by water medium are measured and recorded for distribution and morphology analysis. Significant pressure pulses and temperature gradient almost not occur during MFCI and the molten copper is finely fragmented possibly due to hydrodynamic and thermodynamic effects. And the experimental results are helpful to confirm the prediction of fragmentation mechanism and to validate physical model, which can be applied to the development and validation of analysis code.
{"title":"Experimental Research on Energy Release and Fragments Characteristics Under Molten Materials Discharged Into Liquid Sodium","authors":"Huijun Liang, K. Ge, Yapei Zhang, G. Su, W. Tian, S. Qiu","doi":"10.1115/ICONE26-81400","DOIUrl":"https://doi.org/10.1115/ICONE26-81400","url":null,"abstract":"Hypothetical Core Disruptive Accidents (HCDA) are dominantly concerned during safety assessment and evaluation in Sodium-cooled Fast Reactor (SFR). With molten core materials discharged into liquid sodium, positive reactivity is potentially introduced due to sodium boiling and molten core compaction, which can cause terrible recriticality. The possibility of recriticality and efficient cooling on the relocated debris bed are significantly affected by the fragmentation behavior of molten core in liquid sodium. With few available mechanism models and benchmarks, many investigations have been conducted on the fragmentation characteristics during molten fuel-coolant interaction (MFCI). In the present study, molten copper is used for molten simulant to be discharged into the liquid sodium pool through guiding tube based on a multifunctional experimental facility (COSA). The simulants are heated by electromagnetic induction system in customized ceramic crucible and the molten materials are controlled by magnetic lifting system to be drained through the guiding tube into the bottom liquid sodium pool. Temperature variation and pressure change in the liquid sodium pool are acquired against the energy release during MFCI. Furthermore, the fragments cleaned by water medium are measured and recorded for distribution and morphology analysis. Significant pressure pulses and temperature gradient almost not occur during MFCI and the molten copper is finely fragmented possibly due to hydrodynamic and thermodynamic effects. And the experimental results are helpful to confirm the prediction of fragmentation mechanism and to validate physical model, which can be applied to the development and validation of analysis code.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"56 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"125504210","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}
Quanyao Ren, L. Pan, Wenxiong Zhou, Ting-pu Ye, Hang Liu, Song-song Li
In order to simulate the transfer of mass, momentum and energy in the gas-liquid two-phase flow system, tremendous work focused on the phenomenon, mechanisms and models for two-phase flow in different channels, such as circular pipe, rectangular channel, rod bundle and annulus. Drift-flux model is one of the widely used models for its simplicity and good accuracy, especially for the reactor safety analysis codes (RELAP5 and TRAC et al.) and sub-channel analysis code (COBRA, SILFEED and NASCA et al.). Most of the adopted drift-flux models in these codes were developed based on the void fraction measured in pipe and annulus, which were different with the actual nuclear reactor. Although some drift-flux models were developed for rod bundles, they were based on the void fraction on the whole cross-section not in subchannel in rod bundles due to the lack of effective measuring methods. A novel sub-channel impedance void meter (SCIVM) has been developed to measure the void fraction in sub-channel of 5 × 5 rod bundles, which is adopted to evaluate these existing drift-flux models for rod bundles. By comparison, the values of drift-flux parameters have large differences among different correlations, which are suggested to be reconsidered. Based on the experimental data and physical laws, Lellouche-Zolotar and Chexal-Lellouche correlations show a better performance for drift velocity. If the predicting error of void fraction is the only concerned parameter, Chen-Liu, Ishizuka-Inoue and Chexal-Lellouche correlations are recommended for averaged relative error less than 30%. More experiments are suggested to focus on the distribution parameter and drift velocity through their definition.
{"title":"Comparison of Drift-Flux Models for Void Fraction Prediction in Sub-Channel of Vertical Rod Bundles","authors":"Quanyao Ren, L. Pan, Wenxiong Zhou, Ting-pu Ye, Hang Liu, Song-song Li","doi":"10.1115/ICONE26-81435","DOIUrl":"https://doi.org/10.1115/ICONE26-81435","url":null,"abstract":"In order to simulate the transfer of mass, momentum and energy in the gas-liquid two-phase flow system, tremendous work focused on the phenomenon, mechanisms and models for two-phase flow in different channels, such as circular pipe, rectangular channel, rod bundle and annulus. Drift-flux model is one of the widely used models for its simplicity and good accuracy, especially for the reactor safety analysis codes (RELAP5 and TRAC et al.) and sub-channel analysis code (COBRA, SILFEED and NASCA et al.). Most of the adopted drift-flux models in these codes were developed based on the void fraction measured in pipe and annulus, which were different with the actual nuclear reactor. Although some drift-flux models were developed for rod bundles, they were based on the void fraction on the whole cross-section not in subchannel in rod bundles due to the lack of effective measuring methods. A novel sub-channel impedance void meter (SCIVM) has been developed to measure the void fraction in sub-channel of 5 × 5 rod bundles, which is adopted to evaluate these existing drift-flux models for rod bundles. By comparison, the values of drift-flux parameters have large differences among different correlations, which are suggested to be reconsidered. Based on the experimental data and physical laws, Lellouche-Zolotar and Chexal-Lellouche correlations show a better performance for drift velocity. If the predicting error of void fraction is the only concerned parameter, Chen-Liu, Ishizuka-Inoue and Chexal-Lellouche correlations are recommended for averaged relative error less than 30%. More experiments are suggested to focus on the distribution parameter and drift velocity through their definition.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"6 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"115103615","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Drescher, Brandon De Luna, Marjolein E. Pasman, D. Haas, S. Landsberger
Laboratories in traditional radiochemistry courses typically involve basic and fundamental understanding in solvent extraction, ion exchange, precipitation, etc. procedures. With the increased focus on nuclear forensics in pre- and post-detonation scenarios different skill sets are now required for the student to learn. At the University of Texas we have developed two independent graduate courses in gamma-ray spectrometry and radiochemistry. Currently, we have amalgamated these two courses to 1. better serve our nuclear engineering graduate students, many of which are involved in nuclear forensics and 2. to attract both undergraduate and graduate students from the Chemistry Department. We incorporated gamma-ray spectrometry laboratories with several others which are nuclear forensics related. The seven laboratory sessions include half-life measurement of 137mBa (the daughter produce of 137Cs) and secular equilibrium, basic gamma and beta shielding, and gamma-ray spectrometry calibration, resolution and uncertainty in statistics. These labs have been augmented with four others including uranium fission product identification, 137Cs soil profile with Compton suppression, tritium analysis in water with a liquid scintillation counter and double replacement reaction.
{"title":"Revamping of a Graduate Radiochemistry Course for Nuclear Forensics Applications","authors":"A. Drescher, Brandon De Luna, Marjolein E. Pasman, D. Haas, S. Landsberger","doi":"10.1115/ICONE26-81593","DOIUrl":"https://doi.org/10.1115/ICONE26-81593","url":null,"abstract":"Laboratories in traditional radiochemistry courses typically involve basic and fundamental understanding in solvent extraction, ion exchange, precipitation, etc. procedures. With the increased focus on nuclear forensics in pre- and post-detonation scenarios different skill sets are now required for the student to learn. At the University of Texas we have developed two independent graduate courses in gamma-ray spectrometry and radiochemistry. Currently, we have amalgamated these two courses to 1. better serve our nuclear engineering graduate students, many of which are involved in nuclear forensics and 2. to attract both undergraduate and graduate students from the Chemistry Department. We incorporated gamma-ray spectrometry laboratories with several others which are nuclear forensics related. The seven laboratory sessions include half-life measurement of 137mBa (the daughter produce of 137Cs) and secular equilibrium, basic gamma and beta shielding, and gamma-ray spectrometry calibration, resolution and uncertainty in statistics. These labs have been augmented with four others including uranium fission product identification, 137Cs soil profile with Compton suppression, tritium analysis in water with a liquid scintillation counter and double replacement reaction.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"116 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"126905982","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}
Tao Meng, Sichao Tan, Yuhao He, D. Yuan, Kun Cheng
The space nuclear reactor has been widely studied since 60s in the last century. However, upon the signing of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and the ending of the Cold War, the space nuclear technology gradually faded out the public and researchers view. In recent years, due to the proposal of the United States’ Moon and Mars Project and China’s Deep Space Scientific Project, space nuclear technology are welcoming new opportunities for research and development. In this paper, the Prometheus project design has been analyzed and thereafter been based for multi-kilowatt He-Xe cooled space nuclear system preliminary design. Parameters like system efficiency, compressor ratio, temperature are given and neutron calculations are conducted in order to evaluate its physical performance and provide guidelines for future optimization. A computer program, which can calculate performance of heat pipe radiator, is also coded and thereafter used. At last, some consideration guidelines are concluded for larger power space reactor design.
{"title":"Preliminary Design Considerations of He-Xe Mixture Cooled Space Nuclear Reactor","authors":"Tao Meng, Sichao Tan, Yuhao He, D. Yuan, Kun Cheng","doi":"10.1115/ICONE26-81226","DOIUrl":"https://doi.org/10.1115/ICONE26-81226","url":null,"abstract":"The space nuclear reactor has been widely studied since 60s in the last century. However, upon the signing of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) and the ending of the Cold War, the space nuclear technology gradually faded out the public and researchers view. In recent years, due to the proposal of the United States’ Moon and Mars Project and China’s Deep Space Scientific Project, space nuclear technology are welcoming new opportunities for research and development. In this paper, the Prometheus project design has been analyzed and thereafter been based for multi-kilowatt He-Xe cooled space nuclear system preliminary design. Parameters like system efficiency, compressor ratio, temperature are given and neutron calculations are conducted in order to evaluate its physical performance and provide guidelines for future optimization. A computer program, which can calculate performance of heat pipe radiator, is also coded and thereafter used. At last, some consideration guidelines are concluded for larger power space reactor design.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"3 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124010469","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
As an effective measure to improve the reactor’s inherent safety feature, natural circulation is widely used in current integrated reactor design. The thermal-hydraulic performance of a flashing-driven nature circulation integrated pressurized water reactor (NC-IPWR) is studied, by taking IP100 reactor as reference. The simulation model of the reactor is established by RELAP5 code. A control system is designed based on the operation characteristics of the reactor. Both steady-state and dynamic performance of the reactor are analyzed and the rationality of the control strategy is verified in this work. The results demonstrate the operation characteristics of the IP100 reactor, and the dynamic performance of the reactor during power variation is discussed in detail. The control strategy that keeps the steam pressure and the core outlet temperature constant shows good performance under normal operation conditions. The obtained analysis results are significant for deeper understanding and improving the operation characteristics of the IP100 reactor.
{"title":"Simulation Research on Thermal-Hydraulic Performance of a Natural Circulation Integrated Pressurized Water Reactor","authors":"Yanan Zhao, M. Peng, G. Xia, Lianxin Lv","doi":"10.1115/ICONE26-81059","DOIUrl":"https://doi.org/10.1115/ICONE26-81059","url":null,"abstract":"As an effective measure to improve the reactor’s inherent safety feature, natural circulation is widely used in current integrated reactor design. The thermal-hydraulic performance of a flashing-driven nature circulation integrated pressurized water reactor (NC-IPWR) is studied, by taking IP100 reactor as reference. The simulation model of the reactor is established by RELAP5 code. A control system is designed based on the operation characteristics of the reactor. Both steady-state and dynamic performance of the reactor are analyzed and the rationality of the control strategy is verified in this work. The results demonstrate the operation characteristics of the IP100 reactor, and the dynamic performance of the reactor during power variation is discussed in detail. The control strategy that keeps the steam pressure and the core outlet temperature constant shows good performance under normal operation conditions. The obtained analysis results are significant for deeper understanding and improving the operation characteristics of the IP100 reactor.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"13 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"114340990","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}
Li Zichao, Z. Tao, Shi Shun, Amir Haider, L. Bing, Xiao Zejun
Research on influencing factors of CHF in narrow rectangular channel under natural circulation is of great significance to the safety of reactors. Taking the narrow rectangular experimental device as the research object, influencing factors of CHF in narrow rectangular channel were experimentally studied under natural circulation. With factorial analysis, effects of different factors and their interactions on CHF were analyzed. It is found that the contribution rate of mass flow rate is the largest, followed by the effect of outlet dryness, followed by the effect of system pressure. Their interactions between different factors have little effects on CHF in narrow rectangular channel under natural circulation.
{"title":"Study on Factors Affecting CHF Based on Factorial Analysis in Narrow Rectangular Channel Under Natural Circulation","authors":"Li Zichao, Z. Tao, Shi Shun, Amir Haider, L. Bing, Xiao Zejun","doi":"10.1115/ICONE26-81863","DOIUrl":"https://doi.org/10.1115/ICONE26-81863","url":null,"abstract":"Research on influencing factors of CHF in narrow rectangular channel under natural circulation is of great significance to the safety of reactors. Taking the narrow rectangular experimental device as the research object, influencing factors of CHF in narrow rectangular channel were experimentally studied under natural circulation. With factorial analysis, effects of different factors and their interactions on CHF were analyzed. It is found that the contribution rate of mass flow rate is the largest, followed by the effect of outlet dryness, followed by the effect of system pressure. Their interactions between different factors have little effects on CHF in narrow rectangular channel under natural circulation.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"79 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"124100762","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}
Due to misconceptions surrounding radiation and nuclear energy, educating the general public about basic radiation concepts has become increasingly important. The Virtual Education and Research Laboratory (VERL) at the University of Illinois at Urbana-Champaign (UIUC) has developed a 3D, virtual, interactive game that conveys the physics of radiation and principles of radiation protection to the player. The model is a scavenger hunt style game that takes place in a virtual model of a TRIGA research reactor. Several virtual radiation sources are placed in the 3D virtual model of the TRIGA facility. Radiation drops away from the radiation source. The effect of shielding can also be incorporated in modeling the radiation transport, leading to realistic radiation fields. The user’s goal is to find and collect (virtual) objects placed in this facility while minimizing the dose received in doing so. The player is meant to learn about time, distance, and shielding — key concepts in radiation protection. The start screen displays the radiation field in the form of a colored coded floor, as well as the location of the desired objects. With the given information, the player is encouraged to plan the route to collect the items and minimize exposure. By repeatedly playing the game, the player becomes familiar with the layout of the facility, and of the location of the radiation sources. This educational game is a useful teaching tool. Those unfamiliar with radiation protection concepts are able to understand how important time, distance, and shielding are in minimizing dosage. Additionally, this game proves to be a useful engagement and outreach tool. Upon completion of the game, the user is shown the score, the dose received, as well as a list of dose received in well-known instances such as eating a banana or in getting an x-ray at the dentist’s office. The dose minimization game developed earlier for computers has now been developed for use as a game-app for cell phones. These recent developments allow for wider outreach, further increasing the use of the model as an outreach and educational tool.
{"title":"Dose Minimization Game for Smartphones","authors":"Nolan Stelter, Arnav Das, Zahra Hanifah, R. Uddin","doi":"10.1115/ICONE26-82450","DOIUrl":"https://doi.org/10.1115/ICONE26-82450","url":null,"abstract":"Due to misconceptions surrounding radiation and nuclear energy, educating the general public about basic radiation concepts has become increasingly important. The Virtual Education and Research Laboratory (VERL) at the University of Illinois at Urbana-Champaign (UIUC) has developed a 3D, virtual, interactive game that conveys the physics of radiation and principles of radiation protection to the player. The model is a scavenger hunt style game that takes place in a virtual model of a TRIGA research reactor. Several virtual radiation sources are placed in the 3D virtual model of the TRIGA facility. Radiation drops away from the radiation source. The effect of shielding can also be incorporated in modeling the radiation transport, leading to realistic radiation fields. The user’s goal is to find and collect (virtual) objects placed in this facility while minimizing the dose received in doing so. The player is meant to learn about time, distance, and shielding — key concepts in radiation protection. The start screen displays the radiation field in the form of a colored coded floor, as well as the location of the desired objects. With the given information, the player is encouraged to plan the route to collect the items and minimize exposure. By repeatedly playing the game, the player becomes familiar with the layout of the facility, and of the location of the radiation sources. This educational game is a useful teaching tool. Those unfamiliar with radiation protection concepts are able to understand how important time, distance, and shielding are in minimizing dosage. Additionally, this game proves to be a useful engagement and outreach tool. Upon completion of the game, the user is shown the score, the dose received, as well as a list of dose received in well-known instances such as eating a banana or in getting an x-ray at the dentist’s office.\u0000 The dose minimization game developed earlier for computers has now been developed for use as a game-app for cell phones. These recent developments allow for wider outreach, further increasing the use of the model as an outreach and educational tool.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"19 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"131817523","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}
Due to the increase of computing efficiency and power, full-resolution two-phase flow simulations have become a practical research tool for model development and analysis of reactor flows. The expansion of state-of-the-art high performance computing (HPC) facilities allows for the use of direct numerical simulation (DNS) coupled with Interface Tracking Methods (ITM) to perform full resolution simulations. Given adequate spatial and temporal resolution, DNS can resolve all relevant turbulent scales, allowing for the extraction of high quality and detailed turbulent and two-phase flow numerical data for use in model development. While larger scale bubbly flow DNS are becoming ever more affordable, it is still computationally expensive due to the requirements of the spatial discretization. This presents the largest obstacle for future applications of DNS. For this reason, mesh adaptation techniques are sought after to reduce the computational expense of bubbly flow simulations in complex geometries. By fully resolving only the areas of specific interest, the computational costs of DNS can be reduced. Grid refinement can be based on the location of the interface between the two phases, area of greatest turbulent intensity, averaged bulk fluid velocity data, or the prediction of bubble movement. Coupled with an advanced bubble tracking algorithm, the path of individual bubbles moving through the computational domain can be predicted, and the computational mesh refined within the path area. This refinement can create tracks of greater resolution for the bubbles to move through in the domain, while keeping the bulk resolution of the mesh coarser. Through these means, the overall cost of the simulation is reduced, while high quality numerical data is still obtainable. This work outlines the enhancement of existing mesh adaptation algorithms to implement the bubble tracking refinement, and its practical applications to full resolution two-phase flow simulations.
{"title":"Interface Tracking Simulations of Two-Phase Flow Utilizing Adaptive Meshing Capabilities","authors":"J. Cambareri, I. Bolotnov","doi":"10.1115/ICONE26-81247","DOIUrl":"https://doi.org/10.1115/ICONE26-81247","url":null,"abstract":"Due to the increase of computing efficiency and power, full-resolution two-phase flow simulations have become a practical research tool for model development and analysis of reactor flows. The expansion of state-of-the-art high performance computing (HPC) facilities allows for the use of direct numerical simulation (DNS) coupled with Interface Tracking Methods (ITM) to perform full resolution simulations. Given adequate spatial and temporal resolution, DNS can resolve all relevant turbulent scales, allowing for the extraction of high quality and detailed turbulent and two-phase flow numerical data for use in model development. While larger scale bubbly flow DNS are becoming ever more affordable, it is still computationally expensive due to the requirements of the spatial discretization. This presents the largest obstacle for future applications of DNS.\u0000 For this reason, mesh adaptation techniques are sought after to reduce the computational expense of bubbly flow simulations in complex geometries. By fully resolving only the areas of specific interest, the computational costs of DNS can be reduced. Grid refinement can be based on the location of the interface between the two phases, area of greatest turbulent intensity, averaged bulk fluid velocity data, or the prediction of bubble movement. Coupled with an advanced bubble tracking algorithm, the path of individual bubbles moving through the computational domain can be predicted, and the computational mesh refined within the path area. This refinement can create tracks of greater resolution for the bubbles to move through in the domain, while keeping the bulk resolution of the mesh coarser. Through these means, the overall cost of the simulation is reduced, while high quality numerical data is still obtainable. This work outlines the enhancement of existing mesh adaptation algorithms to implement the bubble tracking refinement, and its practical applications to full resolution two-phase flow simulations.","PeriodicalId":289940,"journal":{"name":"Volume 9: Student Paper Competition","volume":"2020 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2018-07-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"133558732","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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