THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019最新文献
Operator support systems are extensively studied and developed to support human operators for their activities in especially an abnormal condition of a nuclear power plant. By the advancement of computer technology and artificial intelligence, an operator support system can provide detailed support information based on detailed models and utilizing detailed simulation of plant dynamics and/or complicated inference algorithms. However, human operators may not understand the detailed support information under high mental pressure in an abnormal plant condition. In such a case, it is important how to provide essential and understandable support information. This paper deals with a technique to simplify functional models in order to display operator support information that is generated based on detailed functional models. This paper defines eight cognitive states of human operators from the viewpoint of cognitive abilities of human. In addition, three ways to simplify functional models are identified.Operator support systems are extensively studied and developed to support human operators for their activities in especially an abnormal condition of a nuclear power plant. By the advancement of computer technology and artificial intelligence, an operator support system can provide detailed support information based on detailed models and utilizing detailed simulation of plant dynamics and/or complicated inference algorithms. However, human operators may not understand the detailed support information under high mental pressure in an abnormal plant condition. In such a case, it is important how to provide essential and understandable support information. This paper deals with a technique to simplify functional models in order to display operator support information that is generated based on detailed functional models. This paper defines eight cognitive states of human operators from the viewpoint of cognitive abilities of human. In addition, three ways to simplify functional models are identified.
{"title":"A consideration to display operator support information to human operators under high mental pressure","authors":"A. Gofuku","doi":"10.1063/1.5135537","DOIUrl":"https://doi.org/10.1063/1.5135537","url":null,"abstract":"Operator support systems are extensively studied and developed to support human operators for their activities in especially an abnormal condition of a nuclear power plant. By the advancement of computer technology and artificial intelligence, an operator support system can provide detailed support information based on detailed models and utilizing detailed simulation of plant dynamics and/or complicated inference algorithms. However, human operators may not understand the detailed support information under high mental pressure in an abnormal plant condition. In such a case, it is important how to provide essential and understandable support information. This paper deals with a technique to simplify functional models in order to display operator support information that is generated based on detailed functional models. This paper defines eight cognitive states of human operators from the viewpoint of cognitive abilities of human. In addition, three ways to simplify functional models are identified.Operator support systems are extensively studied and developed to support human operators for their activities in especially an abnormal condition of a nuclear power plant. By the advancement of computer technology and artificial intelligence, an operator support system can provide detailed support information based on detailed models and utilizing detailed simulation of plant dynamics and/or complicated inference algorithms. However, human operators may not understand the detailed support information under high mental pressure in an abnormal plant condition. In such a case, it is important how to provide essential and understandable support information. This paper deals with a technique to simplify functional models in order to display operator support information that is generated based on detailed functional models. This paper defines eight cognitive states of human operators from the viewpoint of cognitive abilities of human. In addition, three ways to simplify functional models are identified.","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"5 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74932909","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}
{"title":"In vivo anti-inflammatory activity of anti-atherosclerotic herbs using white male rats (Rattus norvegicus)","authors":"D. Tristantini, R. Amalia","doi":"10.1063/1.5139348","DOIUrl":"https://doi.org/10.1063/1.5139348","url":null,"abstract":"","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"299 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74972186","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}
I. Husnayani, M. Setiawan, P. M. Udiyani, S. Kuntjoro
Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with nominal power of 30 MW and currently utilized for material irradiation and other research purposes. When a sample material is put in the core of RSG-GAS, there will be some amount of nuclear heating generated in the sample material induced by interaction of gamma rays with the sample material. Evaluating the nuclear heating is one of the important aspects regarding the safety of reactor operation and the safety of the sample material itself. In this work, the nuclear heating of several sample materials commonly irradiated in the RSG-GAS core were evaluated using GAMSET code. The sample materials taken as the case study is sample for radioisotope production (TeO2, MoO3, UO2, Sm2O3, Yb2O3, Zn, S), sample for research purpose (C, AlMg3, Hg), topaz, and sample for cladding material (Al, Zr, Fe, SS304L). The sample materials were irradiated in 3 positions in the core, i.e. E6, D9, and B1, for 5 days with thermal power of 15 MW. From the results of nuclear heating calculation, it was found that the nuclear heating generated in sample material in certain position is greatly determined by the type of core structure that surrounding the material position. The difference of nuclear heating generated in the position of D9 has a higher amount of 5% compared to the nuclear heating generated in the position of E6, while for the position of B1 the amount of nuclear heating generated is much lower. Among all the material samples, UO2 has the highest nuclear heating since it contains fissile material, white for the other sample material the amount of nuclear heating varied between between 3∼11 watts/gram in the E6 and D9 position, and between 0.4∼1.4 watts/gram in the B1 position. The results of nuclear heating obtained in this work can be used as a database for the purpose of evaluating the safety of reactor operation and sample material irradiated in RSG-GAS. The data of the nuclear heating in this work can also be used to complement the RSG-GAS safety analysis report.Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with nominal power of 30 MW and currently utilized for material irradiation and other research purposes. When a sample material is put in the core of RSG-GAS, there will be some amount of nuclear heating generated in the sample material induced by interaction of gamma rays with the sample material. Evaluating the nuclear heating is one of the important aspects regarding the safety of reactor operation and the safety of the sample material itself. In this work, the nuclear heating of several sample materials commonly irradiated in the RSG-GAS core were evaluated using GAMSET code. The sample materials taken as the case study is sample for radioisotope production (TeO2, MoO3, UO2, Sm2O3, Yb2O3, Zn, S), sample for research purpose (C, AlMg3, Hg), topaz, and sample for cladding material (Al, Zr, Fe, SS304L). The sampl
{"title":"Evaluation of nuclear heating in sample materials irradiated in RSG – GAS core","authors":"I. Husnayani, M. Setiawan, P. M. Udiyani, S. Kuntjoro","doi":"10.1063/1.5135520","DOIUrl":"https://doi.org/10.1063/1.5135520","url":null,"abstract":"Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with nominal power of 30 MW and currently utilized for material irradiation and other research purposes. When a sample material is put in the core of RSG-GAS, there will be some amount of nuclear heating generated in the sample material induced by interaction of gamma rays with the sample material. Evaluating the nuclear heating is one of the important aspects regarding the safety of reactor operation and the safety of the sample material itself. In this work, the nuclear heating of several sample materials commonly irradiated in the RSG-GAS core were evaluated using GAMSET code. The sample materials taken as the case study is sample for radioisotope production (TeO2, MoO3, UO2, Sm2O3, Yb2O3, Zn, S), sample for research purpose (C, AlMg3, Hg), topaz, and sample for cladding material (Al, Zr, Fe, SS304L). The sample materials were irradiated in 3 positions in the core, i.e. E6, D9, and B1, for 5 days with thermal power of 15 MW. From the results of nuclear heating calculation, it was found that the nuclear heating generated in sample material in certain position is greatly determined by the type of core structure that surrounding the material position. The difference of nuclear heating generated in the position of D9 has a higher amount of 5% compared to the nuclear heating generated in the position of E6, while for the position of B1 the amount of nuclear heating generated is much lower. Among all the material samples, UO2 has the highest nuclear heating since it contains fissile material, white for the other sample material the amount of nuclear heating varied between between 3∼11 watts/gram in the E6 and D9 position, and between 0.4∼1.4 watts/gram in the B1 position. The results of nuclear heating obtained in this work can be used as a database for the purpose of evaluating the safety of reactor operation and sample material irradiated in RSG-GAS. The data of the nuclear heating in this work can also be used to complement the RSG-GAS safety analysis report.Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with nominal power of 30 MW and currently utilized for material irradiation and other research purposes. When a sample material is put in the core of RSG-GAS, there will be some amount of nuclear heating generated in the sample material induced by interaction of gamma rays with the sample material. Evaluating the nuclear heating is one of the important aspects regarding the safety of reactor operation and the safety of the sample material itself. In this work, the nuclear heating of several sample materials commonly irradiated in the RSG-GAS core were evaluated using GAMSET code. The sample materials taken as the case study is sample for radioisotope production (TeO2, MoO3, UO2, Sm2O3, Yb2O3, Zn, S), sample for research purpose (C, AlMg3, Hg), topaz, and sample for cladding material (Al, Zr, Fe, SS304L). The sampl","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"26 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75339286","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}
Iman Kuntoro, S. Pinem, T. M. Sembiring, T. Surbakti
The in-core fuel management of a nuclear power plant is a problem of optimization of core parameters such as operation cycle, average fuel burnup and shut down margin for determining a fuel loading pattern to meet the safety and economic aspects. The study is aimed to obtain an optimal fuel loading pattern. Two models of fuel burn up calculations were taken namely equilibrium and transition cores burn up models. The calculations will be carried out by means of computer codes SRAC2006 for cell calculation and PWR-FUEL for the fuel management. The results of keff values at BOC and EOC for each transition core are approximately 1.05 as the input data and the core cycle length is found to be 330 days. The keff values at both BOC and EOC are very near to critical at equilibrium core and the core cycle length is found 360 days. The results of the calculation of neutron flux distribution and power density using the NODAL and FDM methods of the PWR-FUEL the code has the same results. From the results of the neutronic parameter, it is shown that the optimal loading pattern of PWR core can be determined by the PWR-FUEL code either with equilibrium core search or with transition core burnup models. Key words: fuel loading pattern, PWR-FUEL code, operation safety.The in-core fuel management of a nuclear power plant is a problem of optimization of core parameters such as operation cycle, average fuel burnup and shut down margin for determining a fuel loading pattern to meet the safety and economic aspects. The study is aimed to obtain an optimal fuel loading pattern. Two models of fuel burn up calculations were taken namely equilibrium and transition cores burn up models. The calculations will be carried out by means of computer codes SRAC2006 for cell calculation and PWR-FUEL for the fuel management. The results of keff values at BOC and EOC for each transition core are approximately 1.05 as the input data and the core cycle length is found to be 330 days. The keff values at both BOC and EOC are very near to critical at equilibrium core and the core cycle length is found 360 days. The results of the calculation of neutron flux distribution and power density using the NODAL and FDM methods of the PWR-FUEL the code has the same results. From the results of the neutr...
{"title":"Evaluation of fuel loading pattern of PWR core using PWR-FUEL code","authors":"Iman Kuntoro, S. Pinem, T. M. Sembiring, T. Surbakti","doi":"10.1063/1.5135516","DOIUrl":"https://doi.org/10.1063/1.5135516","url":null,"abstract":"The in-core fuel management of a nuclear power plant is a problem of optimization of core parameters such as operation cycle, average fuel burnup and shut down margin for determining a fuel loading pattern to meet the safety and economic aspects. The study is aimed to obtain an optimal fuel loading pattern. Two models of fuel burn up calculations were taken namely equilibrium and transition cores burn up models. The calculations will be carried out by means of computer codes SRAC2006 for cell calculation and PWR-FUEL for the fuel management. The results of keff values at BOC and EOC for each transition core are approximately 1.05 as the input data and the core cycle length is found to be 330 days. The keff values at both BOC and EOC are very near to critical at equilibrium core and the core cycle length is found 360 days. The results of the calculation of neutron flux distribution and power density using the NODAL and FDM methods of the PWR-FUEL the code has the same results. From the results of the neutronic parameter, it is shown that the optimal loading pattern of PWR core can be determined by the PWR-FUEL code either with equilibrium core search or with transition core burnup models. Key words: fuel loading pattern, PWR-FUEL code, operation safety.The in-core fuel management of a nuclear power plant is a problem of optimization of core parameters such as operation cycle, average fuel burnup and shut down margin for determining a fuel loading pattern to meet the safety and economic aspects. The study is aimed to obtain an optimal fuel loading pattern. Two models of fuel burn up calculations were taken namely equilibrium and transition cores burn up models. The calculations will be carried out by means of computer codes SRAC2006 for cell calculation and PWR-FUEL for the fuel management. The results of keff values at BOC and EOC for each transition core are approximately 1.05 as the input data and the core cycle length is found to be 330 days. The keff values at both BOC and EOC are very near to critical at equilibrium core and the core cycle length is found 360 days. The results of the calculation of neutron flux distribution and power density using the NODAL and FDM methods of the PWR-FUEL the code has the same results. From the results of the neutr...","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"36 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77584976","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}
Composite resin is one of dental material restoration that used in every dental office nowadays. Dental amalgam restoration had several problems in aesthetic and biocompatibility in oral cavity. To overcome the problem, dental composite resin which has great aesthetic, biocompatibility, physical, and mechanical properties has been developed. Composite resins are filled resin and have high compressive strength, abrasion resistance, ease of application, and high translucency. The objective of this review article is to review about dental composite resin including the composition, polymerization process, classification, and physical properties (water sorption, solubility, and polymerization shrinkage) of dental composite resin. Literature relating to dental composite resin and measurement of several physical properties, research methodologies, and contributing factors are selected and reviewed.
{"title":"Dental composite resin: A review","authors":"Yori Rachmia Riva, S. F. Rahman","doi":"10.1063/1.5139331","DOIUrl":"https://doi.org/10.1063/1.5139331","url":null,"abstract":"Composite resin is one of dental material restoration that used in every dental office nowadays. Dental amalgam restoration had several problems in aesthetic and biocompatibility in oral cavity. To overcome the problem, dental composite resin which has great aesthetic, biocompatibility, physical, and mechanical properties has been developed. Composite resins are filled resin and have high compressive strength, abrasion resistance, ease of application, and high translucency. The objective of this review article is to review about dental composite resin including the composition, polymerization process, classification, and physical properties (water sorption, solubility, and polymerization shrinkage) of dental composite resin. Literature relating to dental composite resin and measurement of several physical properties, research methodologies, and contributing factors are selected and reviewed.","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"34 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76116000","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 Indonesia Nuclear Energy Agency (BATAN) has been being managed to design the so called Experimental Power Reactor (EPR), which is a High Temperature Gas Cooled Reactor (HTGR) type with the thermal power of 10 MW. The purpose of the reactor development is a safely demonstration of a small modular nuclear power plant operation. As part of the detail design document of EPR up to the year 2019, the capability to perform the operation based on the determined safety margin have to be described by simulation of the EPR model. The purpose of this work is to simulate the nuclear steam supply system (NSSS) of the EPR, which can demonstrate the steady-state operation performance of the EPR starting from the heat generation in the pebble bed core up to the steam generation in the steam generator component using the RELAP5. Therefore, a complete model of NSSS should consist of primary system and secondary system, which are connected by a piping component consisting of the cold and hot ducts installed in co-axialed way. The simulation of Nuclear Steam Supply System (NSSS) of EPR using RELAP5 results in the output data, which are in general lower than the EPR thermal design data for 100 % core power. For the 50 % core power, the results require further investigation, especially in the methodology of the simulation to achieve the steady-state condition for more representative output. For the 100 % core power, the model of the NSSS of EPR can be used for a selected transient event involving the secondary system.The Indonesia Nuclear Energy Agency (BATAN) has been being managed to design the so called Experimental Power Reactor (EPR), which is a High Temperature Gas Cooled Reactor (HTGR) type with the thermal power of 10 MW. The purpose of the reactor development is a safely demonstration of a small modular nuclear power plant operation. As part of the detail design document of EPR up to the year 2019, the capability to perform the operation based on the determined safety margin have to be described by simulation of the EPR model. The purpose of this work is to simulate the nuclear steam supply system (NSSS) of the EPR, which can demonstrate the steady-state operation performance of the EPR starting from the heat generation in the pebble bed core up to the steam generation in the steam generator component using the RELAP5. Therefore, a complete model of NSSS should consist of primary system and secondary system, which are connected by a piping component consisting of the cold and hot ducts installed in co-axialed...
{"title":"Simulation of nuclear steam supply system of experimental power reactor","authors":"A. S. Ekariansyah, M. Subekti, S. Widodo","doi":"10.1063/1.5135517","DOIUrl":"https://doi.org/10.1063/1.5135517","url":null,"abstract":"The Indonesia Nuclear Energy Agency (BATAN) has been being managed to design the so called Experimental Power Reactor (EPR), which is a High Temperature Gas Cooled Reactor (HTGR) type with the thermal power of 10 MW. The purpose of the reactor development is a safely demonstration of a small modular nuclear power plant operation. As part of the detail design document of EPR up to the year 2019, the capability to perform the operation based on the determined safety margin have to be described by simulation of the EPR model. The purpose of this work is to simulate the nuclear steam supply system (NSSS) of the EPR, which can demonstrate the steady-state operation performance of the EPR starting from the heat generation in the pebble bed core up to the steam generation in the steam generator component using the RELAP5. Therefore, a complete model of NSSS should consist of primary system and secondary system, which are connected by a piping component consisting of the cold and hot ducts installed in co-axialed way. The simulation of Nuclear Steam Supply System (NSSS) of EPR using RELAP5 results in the output data, which are in general lower than the EPR thermal design data for 100 % core power. For the 50 % core power, the results require further investigation, especially in the methodology of the simulation to achieve the steady-state condition for more representative output. For the 100 % core power, the model of the NSSS of EPR can be used for a selected transient event involving the secondary system.The Indonesia Nuclear Energy Agency (BATAN) has been being managed to design the so called Experimental Power Reactor (EPR), which is a High Temperature Gas Cooled Reactor (HTGR) type with the thermal power of 10 MW. The purpose of the reactor development is a safely demonstration of a small modular nuclear power plant operation. As part of the detail design document of EPR up to the year 2019, the capability to perform the operation based on the determined safety margin have to be described by simulation of the EPR model. The purpose of this work is to simulate the nuclear steam supply system (NSSS) of the EPR, which can demonstrate the steady-state operation performance of the EPR starting from the heat generation in the pebble bed core up to the steam generation in the steam generator component using the RELAP5. Therefore, a complete model of NSSS should consist of primary system and secondary system, which are connected by a piping component consisting of the cold and hot ducts installed in co-axialed...","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"25 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74438130","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}
Separation of silicate from mixture of sodium zirconate and sodium silicate by leaching proces using water has been studied. This research aims are to determine the optimum conditions for temperature and contact time on the leached silicate content from mixture of sodium zirconate and sodium silicate, the reaction rate controller, the reaction rate constant (k) and the value of activation energy (Ea) based on shrinking core models. This experiment performed by leaching of a mix between sodium zirconate and sodium silicate by water with the various temperature of 55, 75 and 95 °C and various contact time of 15 minutes to 75 minutes. The resulted of silicate leached was analyzed by Atomic Absorption Spectroscopy (AAS). The maximum amount of silicate leached from mixture of sodium zirconate and sodium silicate which was carried out at weight ratio of feed to solvent volum of 1: 40 and a stirring speed of 220 rpm was 38% at temperature of 95 °C and the contact time of 60 minutes. The leaching kinetics is controlled by chemical reaction with empirical equation is 1-(1-X)1/3 = k1t and the activation energy (Ea) is 45.098 kJ/Mol.Separation of silicate from mixture of sodium zirconate and sodium silicate by leaching proces using water has been studied. This research aims are to determine the optimum conditions for temperature and contact time on the leached silicate content from mixture of sodium zirconate and sodium silicate, the reaction rate controller, the reaction rate constant (k) and the value of activation energy (Ea) based on shrinking core models. This experiment performed by leaching of a mix between sodium zirconate and sodium silicate by water with the various temperature of 55, 75 and 95 °C and various contact time of 15 minutes to 75 minutes. The resulted of silicate leached was analyzed by Atomic Absorption Spectroscopy (AAS). The maximum amount of silicate leached from mixture of sodium zirconate and sodium silicate which was carried out at weight ratio of feed to solvent volum of 1: 40 and a stirring speed of 220 rpm was 38% at temperature of 95 °C and the contact time of 60 minutes. The leaching kinetics is cont...
{"title":"Separation of silicate from mixture of sodium zirconate and sodium silicate by leaching process using water","authors":"M. Setyadji, Sudibyo, Annisa Widyastuti","doi":"10.1063/1.5135554","DOIUrl":"https://doi.org/10.1063/1.5135554","url":null,"abstract":"Separation of silicate from mixture of sodium zirconate and sodium silicate by leaching proces using water has been studied. This research aims are to determine the optimum conditions for temperature and contact time on the leached silicate content from mixture of sodium zirconate and sodium silicate, the reaction rate controller, the reaction rate constant (k) and the value of activation energy (Ea) based on shrinking core models. This experiment performed by leaching of a mix between sodium zirconate and sodium silicate by water with the various temperature of 55, 75 and 95 °C and various contact time of 15 minutes to 75 minutes. The resulted of silicate leached was analyzed by Atomic Absorption Spectroscopy (AAS). The maximum amount of silicate leached from mixture of sodium zirconate and sodium silicate which was carried out at weight ratio of feed to solvent volum of 1: 40 and a stirring speed of 220 rpm was 38% at temperature of 95 °C and the contact time of 60 minutes. The leaching kinetics is controlled by chemical reaction with empirical equation is 1-(1-X)1/3 = k1t and the activation energy (Ea) is 45.098 kJ/Mol.Separation of silicate from mixture of sodium zirconate and sodium silicate by leaching proces using water has been studied. This research aims are to determine the optimum conditions for temperature and contact time on the leached silicate content from mixture of sodium zirconate and sodium silicate, the reaction rate controller, the reaction rate constant (k) and the value of activation energy (Ea) based on shrinking core models. This experiment performed by leaching of a mix between sodium zirconate and sodium silicate by water with the various temperature of 55, 75 and 95 °C and various contact time of 15 minutes to 75 minutes. The resulted of silicate leached was analyzed by Atomic Absorption Spectroscopy (AAS). The maximum amount of silicate leached from mixture of sodium zirconate and sodium silicate which was carried out at weight ratio of feed to solvent volum of 1: 40 and a stirring speed of 220 rpm was 38% at temperature of 95 °C and the contact time of 60 minutes. The leaching kinetics is cont...","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"6 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75838809","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}
Suwoto, H. Adrial, W. Luthfi, T. Setiadipura, Zuhair
The structural materials of Indonesian Experimental Reactor (RDE) is made from graphite that dominate material used on core structure. So that graphite material is very important role, both as core structure material, reflector and also as a moderator and fuel layer and fuel matrix. In thermal neutron energy range, the neutron scattering collision in moderator material such as graphite influences the neutron cross-section and the resulting energy distribution, so that neutrons will get an increase in energy for excitation in the material. Due to high neutron absorption cross section, boron and its compounds find extensive application in the nuclear industry. In actually it is very difficult to obtain pure uranium or thorium dioxide without any other substance. Usually uranium dioxide or thorium kernel always has impurity material like boron. Boron is one of the materials that has strong neutron absorber, specially for Boron-10. The research starting from modeling of kernel TRISO coated fuel particle, spherical pebble fuel and full core modeling by involving multiple heterogeneity calculations. Boron impurities in the TRISO kernel coated fuel particles was carried out with 27 data varied concentration of boron are 0ppm, 1ppm, 2ppm, 3ppm, 4ppm, 5ppm, 6ppm, 7ppm, 8ppm, 9ppm, 10ppm, 15ppm, 20ppm, 25ppm, 30ppm, 35ppm, 30ppm, 35ppm, 40ppm, 45ppm, 50ppm, 60ppm, 70ppm, 80ppm, 80ppm, 90ppm and 100ppm. All calculation analysis will be done using Monte Carlo MCNP6 with continuous neutron energy cross section taken from ENDF/B-VII file. Investigation of multiplication factor effect due to thermal neutron scattering crossing data S(α,β) for graphite and boron impurities on TRISO UO2 or ThO2 kernel coated fuel particle, spherical pebble fuel and full core calculation will be conducted. The all calculation results of the criticality calculation due to effect of boron impurity for both for UO2 and ThO2 kernel coated fuel particles are clearly showed that there are no significant influences effect on multiplication factor value. While criticality calculations using the S(α,β) option for UO2 and ThO2 kernel fuels give the results of a slightly lower multiplication factor with a maximum percentage difference is below than 1,3% for the calculation of the effective multiplication factor on the full core calculation.The structural materials of Indonesian Experimental Reactor (RDE) is made from graphite that dominate material used on core structure. So that graphite material is very important role, both as core structure material, reflector and also as a moderator and fuel layer and fuel matrix. In thermal neutron energy range, the neutron scattering collision in moderator material such as graphite influences the neutron cross-section and the resulting energy distribution, so that neutrons will get an increase in energy for excitation in the material. Due to high neutron absorption cross section, boron and its compounds find extensive application in the nuclear industr
{"title":"Effect of boron impurity and graphite thermal neutron scattering on criticality calculation of Indonesian experimental power reactor","authors":"Suwoto, H. Adrial, W. Luthfi, T. Setiadipura, Zuhair","doi":"10.1063/1.5135511","DOIUrl":"https://doi.org/10.1063/1.5135511","url":null,"abstract":"The structural materials of Indonesian Experimental Reactor (RDE) is made from graphite that dominate material used on core structure. So that graphite material is very important role, both as core structure material, reflector and also as a moderator and fuel layer and fuel matrix. In thermal neutron energy range, the neutron scattering collision in moderator material such as graphite influences the neutron cross-section and the resulting energy distribution, so that neutrons will get an increase in energy for excitation in the material. Due to high neutron absorption cross section, boron and its compounds find extensive application in the nuclear industry. In actually it is very difficult to obtain pure uranium or thorium dioxide without any other substance. Usually uranium dioxide or thorium kernel always has impurity material like boron. Boron is one of the materials that has strong neutron absorber, specially for Boron-10. The research starting from modeling of kernel TRISO coated fuel particle, spherical pebble fuel and full core modeling by involving multiple heterogeneity calculations. Boron impurities in the TRISO kernel coated fuel particles was carried out with 27 data varied concentration of boron are 0ppm, 1ppm, 2ppm, 3ppm, 4ppm, 5ppm, 6ppm, 7ppm, 8ppm, 9ppm, 10ppm, 15ppm, 20ppm, 25ppm, 30ppm, 35ppm, 30ppm, 35ppm, 40ppm, 45ppm, 50ppm, 60ppm, 70ppm, 80ppm, 80ppm, 90ppm and 100ppm. All calculation analysis will be done using Monte Carlo MCNP6 with continuous neutron energy cross section taken from ENDF/B-VII file. Investigation of multiplication factor effect due to thermal neutron scattering crossing data S(α,β) for graphite and boron impurities on TRISO UO2 or ThO2 kernel coated fuel particle, spherical pebble fuel and full core calculation will be conducted. The all calculation results of the criticality calculation due to effect of boron impurity for both for UO2 and ThO2 kernel coated fuel particles are clearly showed that there are no significant influences effect on multiplication factor value. While criticality calculations using the S(α,β) option for UO2 and ThO2 kernel fuels give the results of a slightly lower multiplication factor with a maximum percentage difference is below than 1,3% for the calculation of the effective multiplication factor on the full core calculation.The structural materials of Indonesian Experimental Reactor (RDE) is made from graphite that dominate material used on core structure. So that graphite material is very important role, both as core structure material, reflector and also as a moderator and fuel layer and fuel matrix. In thermal neutron energy range, the neutron scattering collision in moderator material such as graphite influences the neutron cross-section and the resulting energy distribution, so that neutrons will get an increase in energy for excitation in the material. Due to high neutron absorption cross section, boron and its compounds find extensive application in the nuclear industr","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"22 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72842065","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}
Dual fuel (natural gas (NG)-diesel oil (DO)) is one of the interesting strategies that can be applied to diesel engines related to emission reduction. On the other hand its application causes a decrease in diesel engine performance. So it is necessary to look for system settings that can produce the best performance. One possible setting to analyze is the NG injection pulse width. In this study, a variation of NG injection pulse width was carried out to determine the effect on the engine performance at low to high loads. From the experimental results, it can be seen that the variation of gas injection pulse width does not significantly affect torque, power, and BMEP. There is an increase by reducing injection pulse width up to 9 ms in medium and high loads but the changes are small. Variation of injection pulse width 11 ms has the lowest SFOC and the highest thermal efficiency at medium to high loads. While the biggest substitution is obtained by injection pulse width 12 ms but at 75% load, the substitution is lower than 11 ms.
{"title":"Effect of natural gas injection pulse width to diesel dual fuel performance","authors":"M. Zaman, D. Rahmatullah, Semin, F. M. Felayati","doi":"10.1063/1.5138278","DOIUrl":"https://doi.org/10.1063/1.5138278","url":null,"abstract":"Dual fuel (natural gas (NG)-diesel oil (DO)) is one of the interesting strategies that can be applied to diesel engines related to emission reduction. On the other hand its application causes a decrease in diesel engine performance. So it is necessary to look for system settings that can produce the best performance. One possible setting to analyze is the NG injection pulse width. In this study, a variation of NG injection pulse width was carried out to determine the effect on the engine performance at low to high loads. From the experimental results, it can be seen that the variation of gas injection pulse width does not significantly affect torque, power, and BMEP. There is an increase by reducing injection pulse width up to 9 ms in medium and high loads but the changes are small. Variation of injection pulse width 11 ms has the lowest SFOC and the highest thermal efficiency at medium to high loads. While the biggest substitution is obtained by injection pulse width 12 ms but at 75% load, the substitution is lower than 11 ms.","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"13 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72855014","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}
Scoliosis is a medical condition in which a person’s spine has a sideways curve. Treatment to reduce the scoliosis depends on the degree of curve, location, and causes. Surgery is commonly recommended by orthopedists for curves with a high progression by installing instruments that consist of pedicle screws, rods, and connectors. However, many cases of failure both in the implant instruments and the interface of bone and pedicle screw were found caused by high corrective force. The bigger Cobb angle directly means the increase of correction force, which acts on bone-implant interface during scoliosis surgery. In this paper, estimation of corrective forces during scoliosis fixation are investigated using Finite-element analysis (FEA). The research is carried out by modeling a normal and a scoliotic spine with specific Cobb 50.43 degrees. The forces are applied in various numbers of pedicles screws that implanted in thoracic spine, i.e single, three and five pairs of screws. It is found in numerical simulation that the total forces that are needed to fix the scoliotic spine are almost equal for five, three, and five pedicle screws in thoracic spine. However, the maximum force for each screw will increase significantly by reducing the number of screws. The biggest correction force for 5 screws is 54.5 N in the apical section, while it is 218 N for single screws. The higher force applied to a pedicle screw, the higher possibility to get failure and to be pulled out from the bone. It is needed to find the optimal number of using pedicle screws based of the working force, stress and implant cost.
{"title":"Biomechanical analysis of correction force and Cobb angle in a simple model of scoliotic spine fixation","authors":"M. Rusli, N. K. Putra, H. Dahlan, R. Sahputra","doi":"10.1063/1.5138352","DOIUrl":"https://doi.org/10.1063/1.5138352","url":null,"abstract":"Scoliosis is a medical condition in which a person’s spine has a sideways curve. Treatment to reduce the scoliosis depends on the degree of curve, location, and causes. Surgery is commonly recommended by orthopedists for curves with a high progression by installing instruments that consist of pedicle screws, rods, and connectors. However, many cases of failure both in the implant instruments and the interface of bone and pedicle screw were found caused by high corrective force. The bigger Cobb angle directly means the increase of correction force, which acts on bone-implant interface during scoliosis surgery. In this paper, estimation of corrective forces during scoliosis fixation are investigated using Finite-element analysis (FEA). The research is carried out by modeling a normal and a scoliotic spine with specific Cobb 50.43 degrees. The forces are applied in various numbers of pedicles screws that implanted in thoracic spine, i.e single, three and five pairs of screws. It is found in numerical simulation that the total forces that are needed to fix the scoliotic spine are almost equal for five, three, and five pedicle screws in thoracic spine. However, the maximum force for each screw will increase significantly by reducing the number of screws. The biggest correction force for 5 screws is 54.5 N in the apical section, while it is 218 N for single screws. The higher force applied to a pedicle screw, the higher possibility to get failure and to be pulled out from the bone. It is needed to find the optimal number of using pedicle screws based of the working force, stress and implant cost.","PeriodicalId":22239,"journal":{"name":"THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019","volume":"28 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-12-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74859240","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 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019
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