J. M. C. Johari, J. Pane, W. Dewayatna, R. Langenati, G. K. Suryaman, A. S. Adhi, Agus Cahyono, G. Rahmadi, Bambang Herutomo, Sunarko, Dedy Priambodo, Sriyana, Suparman
Abstract A project was initiated to assess the sustainability of Indonesia’s planned nuclear energy system using the IAEA INPRO Methodology to develop an awareness of sustainability issues to support nuclear energy development strategic planning and decision making. Accordingly, for a nuclear energy system to be sustainable, fulfilment of current needs should not compromise the ability of future generations to meet theirs. The sustainability of the planned nuclear energy system was assessed based on the basic principles, user requirements and criteria in the areas of economics, infrastructure, waste management, proliferation resistance, physical protection, environment, and safety. The assessments covers the sustainability of a large reactor completed in 2014 to support the Pre-Feasibility Study in Bangka Belitung, and of a small and medium-sized reactor (SMR) currently ongoing to support the study in West Kalimantan. The results indicate that evidence of achieving sustainability exists in many respects, but there remain gaps to be addressed in due time on aspects such as economics, investment climate, availability of technology, and global development on non-renewable resources. A preliminary study on fuel cycle arrangements is being performed in parallel, focussing on collecting data including on fuel cycle services abroad. Collecting data has been the main challenge for a newcomer country that has yet to decide on the exact reactor technology. The paper is intended to evaluate the performance of Indonesia’s nuclear energy program using the INPRO Methodology to address the sustainability of its planned NES, i.e., reactors and fuel cycle facilities, and to evaluate the effectiveness of the efforts to achieve the purpose. Awareness of sustainability issues plays a key role in nuclear energy policy, especially for determining the energy mix by 2040, to achieve the Net Zero Emission policy target by 2060 or earlier.
{"title":"Evaluating the performance of Indonesia’s nuclear energy program using INPRO methodology","authors":"J. M. C. Johari, J. Pane, W. Dewayatna, R. Langenati, G. K. Suryaman, A. S. Adhi, Agus Cahyono, G. Rahmadi, Bambang Herutomo, Sunarko, Dedy Priambodo, Sriyana, Suparman","doi":"10.1515/kern-2022-0099","DOIUrl":"https://doi.org/10.1515/kern-2022-0099","url":null,"abstract":"Abstract A project was initiated to assess the sustainability of Indonesia’s planned nuclear energy system using the IAEA INPRO Methodology to develop an awareness of sustainability issues to support nuclear energy development strategic planning and decision making. Accordingly, for a nuclear energy system to be sustainable, fulfilment of current needs should not compromise the ability of future generations to meet theirs. The sustainability of the planned nuclear energy system was assessed based on the basic principles, user requirements and criteria in the areas of economics, infrastructure, waste management, proliferation resistance, physical protection, environment, and safety. The assessments covers the sustainability of a large reactor completed in 2014 to support the Pre-Feasibility Study in Bangka Belitung, and of a small and medium-sized reactor (SMR) currently ongoing to support the study in West Kalimantan. The results indicate that evidence of achieving sustainability exists in many respects, but there remain gaps to be addressed in due time on aspects such as economics, investment climate, availability of technology, and global development on non-renewable resources. A preliminary study on fuel cycle arrangements is being performed in parallel, focussing on collecting data including on fuel cycle services abroad. Collecting data has been the main challenge for a newcomer country that has yet to decide on the exact reactor technology. The paper is intended to evaluate the performance of Indonesia’s nuclear energy program using the INPRO Methodology to address the sustainability of its planned NES, i.e., reactors and fuel cycle facilities, and to evaluate the effectiveness of the efforts to achieve the purpose. Awareness of sustainability issues plays a key role in nuclear energy policy, especially for determining the energy mix by 2040, to achieve the Net Zero Emission policy target by 2060 or earlier.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"120 1","pages":"326 - 340"},"PeriodicalIF":0.5,"publicationDate":"2023-05-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75834022","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. J. Gaikwad, N. K. Maheshwari, D. Chandrakar, R. B. Solanki, Bhanuprakash, U. K. Paul
Abstract Many countries are considering Small and Modular Reactors as a viable alternative to counter the climate-change/global-warming with a quick deployment of green, carbon free nuclear energy option in the energy mix. Proponents of SMRs claim that these designs rely more on enhanced inherent/engineered safety and passive features with novel concepts. SMRs are being designed to be fabricated at a factory and then transported as ‘modules’ to the sites for installation either as a single module or multiple module plant. There are many variant of SMRs under considerations/design/construction/commissioning/operation states and majority of the, more than 70 odd SMRs are in the design stage. The paper focuses on safety aspects while addressing the fundamental safety requirement that are derived from fundamental safety principles, the acceptance criteria, the expected/envisaged safety targets and not only the economic impact/considerations. The assessment basis for requirements towards safety enhancements and their extent of assurance in the design are highlighted against the claims made. Ensuring SMR safety with respect to the fundamental safety functions will depend on the foreseen/predicted fission product releases, following overheating of the fuel, during the worst/credible accident conditions and likelihood of occurrence of these accidents. Innovations in the development of advanced fuel, deploying passive safety systems, novel concepts in main heat transport system configuration and advanced features in instrumentation can help in realising the goal of ensured enhanced safety in the SMRs, both in preventive and mitigation domains during severe accidents. Enhancements in the acceptance criteria and deterministic and probabilistic safety targets is also expected and may be envisaged. The paper brings out the challenges faced in the design and regulation of the new NPPs, while addressing fundamental safety principles implementation, generic, specific safety issues, and only genuine innovations can ensure and improve the safety. Aspects related to passive systems and the optimal main heat removal system configuration of the NPPs are also discussed. The aspects related to concurrent design and regulation of new NPPs including SMRs also has been brought out in the paper.
{"title":"Ensuring safety of new, advanced small modular reactors for fundamental safety and with an optimal main heat transport systems configuration","authors":"A. J. Gaikwad, N. K. Maheshwari, D. Chandrakar, R. B. Solanki, Bhanuprakash, U. K. Paul","doi":"10.1515/kern-2022-0106","DOIUrl":"https://doi.org/10.1515/kern-2022-0106","url":null,"abstract":"Abstract Many countries are considering Small and Modular Reactors as a viable alternative to counter the climate-change/global-warming with a quick deployment of green, carbon free nuclear energy option in the energy mix. Proponents of SMRs claim that these designs rely more on enhanced inherent/engineered safety and passive features with novel concepts. SMRs are being designed to be fabricated at a factory and then transported as ‘modules’ to the sites for installation either as a single module or multiple module plant. There are many variant of SMRs under considerations/design/construction/commissioning/operation states and majority of the, more than 70 odd SMRs are in the design stage. The paper focuses on safety aspects while addressing the fundamental safety requirement that are derived from fundamental safety principles, the acceptance criteria, the expected/envisaged safety targets and not only the economic impact/considerations. The assessment basis for requirements towards safety enhancements and their extent of assurance in the design are highlighted against the claims made. Ensuring SMR safety with respect to the fundamental safety functions will depend on the foreseen/predicted fission product releases, following overheating of the fuel, during the worst/credible accident conditions and likelihood of occurrence of these accidents. Innovations in the development of advanced fuel, deploying passive safety systems, novel concepts in main heat transport system configuration and advanced features in instrumentation can help in realising the goal of ensured enhanced safety in the SMRs, both in preventive and mitigation domains during severe accidents. Enhancements in the acceptance criteria and deterministic and probabilistic safety targets is also expected and may be envisaged. The paper brings out the challenges faced in the design and regulation of the new NPPs, while addressing fundamental safety principles implementation, generic, specific safety issues, and only genuine innovations can ensure and improve the safety. Aspects related to passive systems and the optimal main heat removal system configuration of the NPPs are also discussed. The aspects related to concurrent design and regulation of new NPPs including SMRs also has been brought out in the paper.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"23 1","pages":"475 - 490"},"PeriodicalIF":0.5,"publicationDate":"2023-05-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79285770","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract Study on calculation model and risk area of radionuclide diffusion in coastal waters under nuclear leakage accidents with different levels can help predict and evaluate consequences of radionuclide leakage accidents. Thus they play an important role in emergency response and accident mitigation. In the first step of the study, a climate hydrodynamic model in China coastal waters was established based on the climate data. In the next step, according to the real-time meteorological data, a hydrodynamic model in coastal waters of Haiyang nuclear power station was founded using the result of the climate hydrodynamic as a boundary. Then, according to the result of the hydrodynamic model in coastal waters of Haiyang nuclear power station, a radionuclide diffusion model in coastal waters of Haiyang nuclear power station was set up, in which the Euler method was adopted. With the radionuclide diffusion model, the total leaked radioactivity of radionuclides was set from 1018 Bq to 1012 Bq with a decrease of every two orders of magnitude. Thus, scenarios of radionuclide diffusion under assumed nuclear leakage accidents with different levels were calculated and their corresponding risk area were analyzed under the assumption that radionuclides leaked for consecutive five days. The results show that when the leaked radioactivity of radionuclides is 1018 Bq, the risk area on the seventh day is about 41 km east, 22 km south and 19 km west of the power station; on the fourteenth day, the risk area is about 65 km east, 22 km south and 25 km west of the power station. When the total leaked radioactivity of radionuclides declines by two orders of magnitude, the risk area will be reduced by about 10 km–20 km in the east direction accordingly. When it declines to 1014 Bq, the risk area decreases sharply to a small area. When it declines to 1012 Bq, the risk area is barely found. This model was verified from two aspects, namely the flow field and the radionuclide concentration. Hydrodynamic results can well describe the Yellow Sea cold water mass, Yellow Sea warm current and tidal current. Changes of radioactivity in different positions are fundamentally consistent with that in Fuikushima nuclear leakage accident. It indicates the hydrodynamic model and radionuclide diffusion model in the study are feasible and reliable.
{"title":"Study on calculation model and risk area of radionuclide diffusion in coastal waters under nuclear leakage accidents with different levels","authors":"Zichao Li, Rong-chang Chen, Tao Zhou, Chen Liu, Guangcheng Si, Qingqing Xue","doi":"10.1515/kern-2022-0120","DOIUrl":"https://doi.org/10.1515/kern-2022-0120","url":null,"abstract":"Abstract Study on calculation model and risk area of radionuclide diffusion in coastal waters under nuclear leakage accidents with different levels can help predict and evaluate consequences of radionuclide leakage accidents. Thus they play an important role in emergency response and accident mitigation. In the first step of the study, a climate hydrodynamic model in China coastal waters was established based on the climate data. In the next step, according to the real-time meteorological data, a hydrodynamic model in coastal waters of Haiyang nuclear power station was founded using the result of the climate hydrodynamic as a boundary. Then, according to the result of the hydrodynamic model in coastal waters of Haiyang nuclear power station, a radionuclide diffusion model in coastal waters of Haiyang nuclear power station was set up, in which the Euler method was adopted. With the radionuclide diffusion model, the total leaked radioactivity of radionuclides was set from 1018 Bq to 1012 Bq with a decrease of every two orders of magnitude. Thus, scenarios of radionuclide diffusion under assumed nuclear leakage accidents with different levels were calculated and their corresponding risk area were analyzed under the assumption that radionuclides leaked for consecutive five days. The results show that when the leaked radioactivity of radionuclides is 1018 Bq, the risk area on the seventh day is about 41 km east, 22 km south and 19 km west of the power station; on the fourteenth day, the risk area is about 65 km east, 22 km south and 25 km west of the power station. When the total leaked radioactivity of radionuclides declines by two orders of magnitude, the risk area will be reduced by about 10 km–20 km in the east direction accordingly. When it declines to 1014 Bq, the risk area decreases sharply to a small area. When it declines to 1012 Bq, the risk area is barely found. This model was verified from two aspects, namely the flow field and the radionuclide concentration. Hydrodynamic results can well describe the Yellow Sea cold water mass, Yellow Sea warm current and tidal current. Changes of radioactivity in different positions are fundamentally consistent with that in Fuikushima nuclear leakage accident. It indicates the hydrodynamic model and radionuclide diffusion model in the study are feasible and reliable.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"19 1","pages":"491 - 502"},"PeriodicalIF":0.5,"publicationDate":"2023-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89415589","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Fitria Miftasani, N. Widiawati, Nuri Trianti, T. Setiadipura, Z. Zuhair, D. Irwanto, S. Permana, Z. Su’ud
Abstract TRISO fuel particle using ZrC has better strength and resistance to high temperatures than SiC. Previous studies show that the ZrC layer, as a substitution of SiC within the TRISO layer of coated fuel particles, has an insignificant difference in the performance of the neutronic aspect. Further neutronic studies are required to obtain the best combination of thorium-based fuel with ZrC coating for HTGR. This study analyzed the neutronic performance of three types of thorium-based fuels, oxide, carbide, and nitride, for HTGR. The reactor design refers to the High-Temperature Test Reactor with some axial and radial fuel configuration adjustments. This reactor is designed to operate at 200 MWt and has been modified to use a ZrC layer as a substitute for the SiC layer on the coated fuel particles. The neutronic study is carried out using SRAC2006 code with JENDL 4.0 nuclear data library. Neutronic parameters analyzed include multiplication factor, power peaking factor, and neutron spectrum. Neutronic analysis results show that thorium nitride fuel’s multiplication factor (keff) is better than other compared fuel types with k-eff 1.050, higher than thorium carbide, 1.004. At the same time, thorium oxide has been sub-critical. The power-peaking value of all materials is close to the ideal peaking value that is one. Other neutronic aspects, such as the neutron spectrum for three compared fuel types, have a similar trend.
{"title":"Optimization of 200 MWt HTGR with ThUN-based fuel and zirconium carbide TRISO layer","authors":"Fitria Miftasani, N. Widiawati, Nuri Trianti, T. Setiadipura, Z. Zuhair, D. Irwanto, S. Permana, Z. Su’ud","doi":"10.1515/kern-2023-0003","DOIUrl":"https://doi.org/10.1515/kern-2023-0003","url":null,"abstract":"Abstract TRISO fuel particle using ZrC has better strength and resistance to high temperatures than SiC. Previous studies show that the ZrC layer, as a substitution of SiC within the TRISO layer of coated fuel particles, has an insignificant difference in the performance of the neutronic aspect. Further neutronic studies are required to obtain the best combination of thorium-based fuel with ZrC coating for HTGR. This study analyzed the neutronic performance of three types of thorium-based fuels, oxide, carbide, and nitride, for HTGR. The reactor design refers to the High-Temperature Test Reactor with some axial and radial fuel configuration adjustments. This reactor is designed to operate at 200 MWt and has been modified to use a ZrC layer as a substitute for the SiC layer on the coated fuel particles. The neutronic study is carried out using SRAC2006 code with JENDL 4.0 nuclear data library. Neutronic parameters analyzed include multiplication factor, power peaking factor, and neutron spectrum. Neutronic analysis results show that thorium nitride fuel’s multiplication factor (keff) is better than other compared fuel types with k-eff 1.050, higher than thorium carbide, 1.004. At the same time, thorium oxide has been sub-critical. The power-peaking value of all materials is close to the ideal peaking value that is one. Other neutronic aspects, such as the neutron spectrum for three compared fuel types, have a similar trend.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"3 1","pages":"503 - 517"},"PeriodicalIF":0.5,"publicationDate":"2023-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85312573","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract In this paper, the time evolution of bremsstrahlung radiation loss, plasma frequency and electron particles density and the relationship between these parameters and black body radiation are investigated. The model used in this work is based on numerical solution of particle and energy balance equations in ITER with DT fuel. The fusion reaction takes places in a plasma of deuterium and tritium heated to millions of degrees. It is expected that at this temperature, the thermal noise could have a significant effect on plasma behavior. This effect is considered in the solution of equations for the first time in this work. In order to attain a proper set of particle and energy balance equations, an appropriate thermal noise term is considered in the set of coupled differential equations. These equations are solved simultaneously by numerical methods. The results of the calculations for bremsstrahlung radiation loss, plasma frequency, intensity of blackbody radiation, absorption coefficient and quality factor show that in the absence of thermal noise blackbody radiation doesn’t occur but in the presence of thermal noise blackbody radiation occurs in times of 55.7 s and 42.73 s for two cases of considering and ignoring impurity respectively. As it can be seen that with the addition of impurities to the system, bremsstrahlung radiation and intensity of blackbody radiation increase while absorption coefficient and quality factor decrease.
{"title":"Role of impurity and thermal noise on the radiation sources in ITER using DT fuel","authors":"R. Khoramdel, S. Hosseinimotlagh, Z. Parang","doi":"10.1515/kern-2023-0005","DOIUrl":"https://doi.org/10.1515/kern-2023-0005","url":null,"abstract":"Abstract In this paper, the time evolution of bremsstrahlung radiation loss, plasma frequency and electron particles density and the relationship between these parameters and black body radiation are investigated. The model used in this work is based on numerical solution of particle and energy balance equations in ITER with DT fuel. The fusion reaction takes places in a plasma of deuterium and tritium heated to millions of degrees. It is expected that at this temperature, the thermal noise could have a significant effect on plasma behavior. This effect is considered in the solution of equations for the first time in this work. In order to attain a proper set of particle and energy balance equations, an appropriate thermal noise term is considered in the set of coupled differential equations. These equations are solved simultaneously by numerical methods. The results of the calculations for bremsstrahlung radiation loss, plasma frequency, intensity of blackbody radiation, absorption coefficient and quality factor show that in the absence of thermal noise blackbody radiation doesn’t occur but in the presence of thermal noise blackbody radiation occurs in times of 55.7 s and 42.73 s for two cases of considering and ignoring impurity respectively. As it can be seen that with the addition of impurities to the system, bremsstrahlung radiation and intensity of blackbody radiation increase while absorption coefficient and quality factor decrease.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"227 1","pages":"446 - 456"},"PeriodicalIF":0.5,"publicationDate":"2023-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74807754","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract The research addressed the economic evaluation of nuclear energy, coal, oil and natural gas in various cycle configurations with different distillation technologies. A comparison between fossil and nuclear energy sources is discussed. A comprehensive review of all desalination plants using fossil and nuclear energy with cogeneration is also included and discussed. Computational model was used to evaluate nuclear and non-nuclear desalination plants. Four possibilities were discussed. In this first case, a nuclear gas cycle integrated with a MED desalination plant was found to have the lowest cost in the gas cycle. While the nuclear steam cycle integrated with RO has the lowest cost compared to oil and natural gas. In the third case, the combined nuclear cycle was discussed and it was found that the combined nuclear cycle associated with RO has the lowest cost. In the last case, the evaluation of a heat-only desalination plant was conducted among all sources, and it was found that the nuclear plant with RO and MED has the lowest cost. In addition, the completed nuclear desalination plant was subjected to five scenarios to calculate and estimate at which capacity the plant provides the best values. The results show that nuclear desalination with gas cycle is the most economical among the oil and natural gas options.
{"title":"Computational analysis of nuclear desalination system under various configurations","authors":"Salah Ud-Din Khan, J. Orfi","doi":"10.1515/kern-2022-0100","DOIUrl":"https://doi.org/10.1515/kern-2022-0100","url":null,"abstract":"Abstract The research addressed the economic evaluation of nuclear energy, coal, oil and natural gas in various cycle configurations with different distillation technologies. A comparison between fossil and nuclear energy sources is discussed. A comprehensive review of all desalination plants using fossil and nuclear energy with cogeneration is also included and discussed. Computational model was used to evaluate nuclear and non-nuclear desalination plants. Four possibilities were discussed. In this first case, a nuclear gas cycle integrated with a MED desalination plant was found to have the lowest cost in the gas cycle. While the nuclear steam cycle integrated with RO has the lowest cost compared to oil and natural gas. In the third case, the combined nuclear cycle was discussed and it was found that the combined nuclear cycle associated with RO has the lowest cost. In the last case, the evaluation of a heat-only desalination plant was conducted among all sources, and it was found that the nuclear plant with RO and MED has the lowest cost. In addition, the completed nuclear desalination plant was subjected to five scenarios to calculate and estimate at which capacity the plant provides the best values. The results show that nuclear desalination with gas cycle is the most economical among the oil and natural gas options.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"26 1","pages":"291 - 301"},"PeriodicalIF":0.5,"publicationDate":"2023-04-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80967890","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
H. Zaky, Mohamed G Abd Elfatah, S. El-Mongy, M. Abdel‐Rahman
Abstract Mobile Robots (MR) are currently used across a variety of different sectors and have military, nuclear and industrial applications among others. In unmanned systems, teleoperation sensors, navigation instruments, control systems and radiation sensors can be fixed on the MR to perform required tasks such as radiological scanning, identifying, and surveying the contaminated environment that has been exposed to radiation. In this work, an estimation of the mobile robot location and the optimum path for time-delay compensation for MR teleoperation are investigated. As the MR teleoperation has a stochastic nature, the kinematics equations are modeled using stochastic differential equations (SDEs). Afterwards, these SDEs are solved using Numerical algorithms such as Euler–Maruyama algorithm which is used to approximate SDEs solution with the aid of MATLAB. Additionally, the results are discussed and depicted in tables and figures. Finally, the simulated results for the solution are performed and are found to highly agree with the ideal path of the simulated MR. This result is of great importance to be used in the case of nuclear emergency response and mitigation.
{"title":"Euler–Maruyama algorithm in estimating UGV path and location in nuclear emergency and security applications","authors":"H. Zaky, Mohamed G Abd Elfatah, S. El-Mongy, M. Abdel‐Rahman","doi":"10.1515/kern-2022-0102","DOIUrl":"https://doi.org/10.1515/kern-2022-0102","url":null,"abstract":"Abstract Mobile Robots (MR) are currently used across a variety of different sectors and have military, nuclear and industrial applications among others. In unmanned systems, teleoperation sensors, navigation instruments, control systems and radiation sensors can be fixed on the MR to perform required tasks such as radiological scanning, identifying, and surveying the contaminated environment that has been exposed to radiation. In this work, an estimation of the mobile robot location and the optimum path for time-delay compensation for MR teleoperation are investigated. As the MR teleoperation has a stochastic nature, the kinematics equations are modeled using stochastic differential equations (SDEs). Afterwards, these SDEs are solved using Numerical algorithms such as Euler–Maruyama algorithm which is used to approximate SDEs solution with the aid of MATLAB. Additionally, the results are discussed and depicted in tables and figures. Finally, the simulated results for the solution are performed and are found to highly agree with the ideal path of the simulated MR. This result is of great importance to be used in the case of nuclear emergency response and mitigation.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"341 1","pages":"361 - 369"},"PeriodicalIF":0.5,"publicationDate":"2023-04-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90316593","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract Nuclear desalination has been identified as an alternative option with much lower carbon dioxide emissions to provide fresh water by driving high capacity desalination plants. This work considers a theoretical analysis of using nuclear desalination to provide fresh water in three selected Saudi Arabian cities. It presents a theoretical model that integrates the characteristics of nuclear reactor, power cycle and desalination blocks. The power block includes three steam turbines and five feed water heaters. It is coupled via low pressure turbine to a multiple effect desalination unit integrated with a thermal vapor compressor encompassing eight effects, seven feed preheaters and a down condenser. The output includes work generated as function of fuel mass and reactor type, enrichment percentage, power and water production with different nuclear reactor type. Desalination performance indicators such as the fresh water production rate, gain output ratio (GOR), specific energy consumption (SEC) and specific cooling water mass flow rate have been evaluated and analyzed as function of sea water temperature for three specific Saudi cities. It was found that these indicators reflect better performance along a year for Jizan city than for Jubail and Tabuk. The case of Jizan city gives over the whole year more uniform values of water production rates, gain output ratio, specific energy consumption and cooling water mass flow rates.
{"title":"Performance analysis of nuclear powered desalination unit based on MED-TVC: a case study for Saudi Arabia","authors":"Salah Ud-Din Khan, A. Najib, J. Orfi","doi":"10.1515/kern-2023-0007","DOIUrl":"https://doi.org/10.1515/kern-2023-0007","url":null,"abstract":"Abstract Nuclear desalination has been identified as an alternative option with much lower carbon dioxide emissions to provide fresh water by driving high capacity desalination plants. This work considers a theoretical analysis of using nuclear desalination to provide fresh water in three selected Saudi Arabian cities. It presents a theoretical model that integrates the characteristics of nuclear reactor, power cycle and desalination blocks. The power block includes three steam turbines and five feed water heaters. It is coupled via low pressure turbine to a multiple effect desalination unit integrated with a thermal vapor compressor encompassing eight effects, seven feed preheaters and a down condenser. The output includes work generated as function of fuel mass and reactor type, enrichment percentage, power and water production with different nuclear reactor type. Desalination performance indicators such as the fresh water production rate, gain output ratio (GOR), specific energy consumption (SEC) and specific cooling water mass flow rate have been evaluated and analyzed as function of sea water temperature for three specific Saudi cities. It was found that these indicators reflect better performance along a year for Jizan city than for Jubail and Tabuk. The case of Jizan city gives over the whole year more uniform values of water production rates, gain output ratio, specific energy consumption and cooling water mass flow rates.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"16 1","pages":"302 - 315"},"PeriodicalIF":0.5,"publicationDate":"2023-04-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79699161","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract A correlation for the volume porosity of irradiated UO2 fuel as a sum of the individual contributions of the pore and swelling porosity in terms of the local density and fractional matrix swelling is developed. By using the existing low temperature low burnup model of swelling, evolution of the matrix swelling and porosity terms are calculated for UO2 fuel in the low burnup but in the high burnup two various models are applied, the one with considering grain recrystallization and another without it but with the method of Xe depletion measurement. The purpose of the paper is comparison of fuel swelling behavior between two models at high burnups. The bulk swelling and porosity evolution in both methods are also validated by experimental data. Finally, one conclusion from this comparison is obtained, as the method which considering grain recrystallization has more rational behavior in the fuel swelling and porosity.
{"title":"Comparative analysis of swelling and porosity evolution in UO2 fuel via two approaches","authors":"B. Roostaii","doi":"10.1515/kern-2022-0121","DOIUrl":"https://doi.org/10.1515/kern-2022-0121","url":null,"abstract":"Abstract A correlation for the volume porosity of irradiated UO2 fuel as a sum of the individual contributions of the pore and swelling porosity in terms of the local density and fractional matrix swelling is developed. By using the existing low temperature low burnup model of swelling, evolution of the matrix swelling and porosity terms are calculated for UO2 fuel in the low burnup but in the high burnup two various models are applied, the one with considering grain recrystallization and another without it but with the method of Xe depletion measurement. The purpose of the paper is comparison of fuel swelling behavior between two models at high burnups. The bulk swelling and porosity evolution in both methods are also validated by experimental data. Finally, one conclusion from this comparison is obtained, as the method which considering grain recrystallization has more rational behavior in the fuel swelling and porosity.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"47 1","pages":"354 - 360"},"PeriodicalIF":0.5,"publicationDate":"2023-04-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80602810","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abuzar Shakeri, E. Heidari, Nasrin Hosseini Motlagh, H. Vanaie
Abstract In heavy ion fusion using the inertial confinement fusion (ICF) approach, firstly, the deposited energy of heavy ions in the target and, secondly, the charged products resulting from fusion reactions in the plasma of the fuel capsule are key and necessary points. In this paper, we used the ICF method for the core of a spherical fusion reactor simulation filled with multi-layer fuel capsules with foam using symmetrical irradiation from 32 different directions by two heavy ion beams of Cs and Pb with radiation energies of 8 and 10 GeV, respectively. Then we simulated the process of penetration and deposited energy of the beams inside the core of this reactor using GEANT4 code. The results of our simulations show that if the atomic number of radiation beams increases, the amount of beam stopping power increases, which is in agreement with existing theories. Also, by changing parameters such as the type and energy amount of the radiation beam, thickness, and the type of material selected in the layers of the desired fuel capsules, the amount of the penetration depth, the produced secondary particles, the stopping power per unit volume of fuel capsule and the reactor core will change. Eventually, these variations will cause a change in deposited energy gain inside the core of a spherical fusion reactor. The obtained maximum deposited energy due to the two selective Pb+ and Cs+ beams with 8 and 10 GeV energies in this study is related to DT fuel compared to the two neutron free-fuels of D3He and P11B. It can be seen that energy gain increases significantly with changing beam energy from 8 to 10 GeV, but for both selected energy, the enhancement of DT energy gain compared to D3He and P11B is not so significant.
{"title":"Modeling and simulation of deposited energy gain via irradiation of heavy ion beams on the fusion reactor contains spherical fuel capsules with foam","authors":"Abuzar Shakeri, E. Heidari, Nasrin Hosseini Motlagh, H. Vanaie","doi":"10.1515/kern-2022-0098","DOIUrl":"https://doi.org/10.1515/kern-2022-0098","url":null,"abstract":"Abstract In heavy ion fusion using the inertial confinement fusion (ICF) approach, firstly, the deposited energy of heavy ions in the target and, secondly, the charged products resulting from fusion reactions in the plasma of the fuel capsule are key and necessary points. In this paper, we used the ICF method for the core of a spherical fusion reactor simulation filled with multi-layer fuel capsules with foam using symmetrical irradiation from 32 different directions by two heavy ion beams of Cs and Pb with radiation energies of 8 and 10 GeV, respectively. Then we simulated the process of penetration and deposited energy of the beams inside the core of this reactor using GEANT4 code. The results of our simulations show that if the atomic number of radiation beams increases, the amount of beam stopping power increases, which is in agreement with existing theories. Also, by changing parameters such as the type and energy amount of the radiation beam, thickness, and the type of material selected in the layers of the desired fuel capsules, the amount of the penetration depth, the produced secondary particles, the stopping power per unit volume of fuel capsule and the reactor core will change. Eventually, these variations will cause a change in deposited energy gain inside the core of a spherical fusion reactor. The obtained maximum deposited energy due to the two selective Pb+ and Cs+ beams with 8 and 10 GeV energies in this study is related to DT fuel compared to the two neutron free-fuels of D3He and P11B. It can be seen that energy gain increases significantly with changing beam energy from 8 to 10 GeV, but for both selected energy, the enhancement of DT energy gain compared to D3He and P11B is not so significant.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"23 1","pages":"370 - 382"},"PeriodicalIF":0.5,"publicationDate":"2023-04-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82142124","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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